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1

Calculation notes in support of TWRS FSAR spray leak accident analysis  

SciTech Connect

This document contains the detailed calculations that support the spray leak accident analysis in the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR). The consequence analyses in this document form the basis for the selection of controls to mitigate or prevent spray leaks throughout TWRS. Pressurized spray leaks can occur due to a breach in containment barriers along transfer routes, during waste transfers. Spray leaks are of particular safety concern because, depending on leak dimensions, and waste pressure, they can be relatively efficient generators of dispersible sized aerosols that can transport downwind to onsite and offsite receptors. Waste is transferred between storage tanks and between processing facilities and storage tanks in TWRS through a system of buried transfer lines. Pumps for transferring waste and jumpers and valves for rerouting waste are located inside below grade pits and structures that are normally covered. Pressurized spray leaks can emanate to the atmosphere due to breaches in waste transfer associated equipment inside these structures should the structures be uncovered at the time of the leak. Pressurized spray leaks can develop through holes or cracks in transfer piping, valve bodies or pump casings caused by such mechanisms as corrosion, erosion, thermal stress, or water hammer. Leaks through degraded valve packing, jumper gaskets, or pump seals can also result in pressurized spray releases. Mechanisms that can degrade seals, packing and gaskets include aging, radiation hardening, thermal stress, etc. An1782other common cause for spray leaks inside transfer enclosures are misaligned jumpers caused by human error. A spray leak inside a DST valve pit during a transfer of aging waste was selected as the bounding, representative accident for detailed analysis. Sections 2 through 5 below develop this representative accident using the DOE- STD-3009 format. Sections 2 describes the unmitigated and mitigated accident scenarios evaluated to determine the need for safety class SSCs or TSR controls. Section 3 develops the source terms associated with the unmitigated and mitigated accident scenarios. Section 4 estimates the radiological and toxicological consequences for the unmitigated and mitigated scenarios. Section 5 compares the radiological and toxicological consequences against the TWRS evaluation guidelines. Section 6 extrapolates from the representative accident case to other represented spray leak sites to assess the conservatism in using the representative case to define controls for other postulated spray leak sites throughout TWRS. Section 7 discusses the sensitivities of the consequence analyses to the key parameters and assumptions used in the analyses. Conclusions are drawn in Section 8. The analyses herein pertain to spray leaks initiated due to internal mechanisms (e.g., corrosion, erosion, thermal stress, etc). External initiators of spray leaks (e.g., excavation accidents), and natural phenomena initiators (e.g., seismic events) are to be covered in separate accident analyses.

Hall, B.W.

1996-09-25T23:59:59.000Z

2

Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR  

Science Conference Proceedings (OSTI)

The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

Sebrell, W.

1983-07-01T23:59:59.000Z

3

Risks From Severe Accidents Involving Steam Generator Tube Leaks or Ruptures  

Science Conference Proceedings (OSTI)

The various types of corrosion observed in PWR steam generator tubes prompted the nuclear industry to initiate a program of Steam Generator Degradation Specific Management (SGDSM). This program's objective is to develop a cost-effective means to maintain plant safety while improving steam generator reliability. Critical to this program is an assessment of the impact of steam generator tube leakage or rupture during severe accidents. This study determined the contributions of these types of severe acciden...

1998-01-02T23:59:59.000Z

4

Accidents  

NLE Websites -- All DOE Office Websites (Extended Search)

Health Risks » Accidents Health Risks » Accidents DUF6 Health Risks line line Accidents Storage Conversion Manufacturing Disposal Transportation Accidents A discussion of accidents involving depleted UF6 storage cylinders, including possible health effects, accident risk, and accident history. Potential Health Effects from Cylinder Accidents Accidents involving depleted UF6 storage cylinders are a concern because they could result in an uncontrolled release of UF6 to the environment, which could potentially affect the health of workers and members of the public living downwind of the accident site. Accidental release of UF6 from storage cylinders or during processing activities could result in injuries or fatalities. The most immediate hazard after a release would be from inhalation of hydrogen fluoride (HF), a highly corrosive gas formed when

5

Leak Pruning  

E-Print Network (OSTI)

Managed languages improve programmer productivity with type safety and garbage collection, which eliminate memory errors such as dangling pointers, double frees, and buffer overflows. However, programs may still leak memory if programmers forget to eliminate the last reference to an object that will not be used again. Leaks slow programs by increasing collector workload and frequency. Growing leaks crash programs. Instead of crashing, leak pruning extends program availability by predicting and reclaiming leaked objects at run time. Whereas garbage collection over-approximates live objects using reachability, leak pruning predicts dead objects and reclaims them based on how stale they are and the size of stale data structures. Leak pruning preserves semantics because it waits for heap exhaustion before reclaiming objects and then poisons references to objects it reclaims. If the program later tries to access these objects, the virtual machine (VM) throws an internal error. We implement leak pruning in a Java VM, show its overhead is low, and evaluate it on 10 leaking programs. Leak pruning does not help two programs, executes four substantial programs 1.6-35X longer, and executes four programs, including two leaks in Eclipse, for at least 24 hours. In the worst case, leak pruning defers fatal errors. In the best case, programs with unbounded memory requirements execute indefinitely and correctly in bounded memory with consistent throughput.

Michael D. Bond; Kathryn S. McKinley

2009-01-01T23:59:59.000Z

6

Hazard Analysis for In Tank Spray Leaks  

SciTech Connect

The River Protection Project (RPP) Authorization Basis (AB) contains controls that address spray leaks in tanks. However, there are no hazardous conditions in the Hazards Database that specifically identify in-tank spray leak scenarios. The purpose of this Hazards Evaluation is to develop hazardous conditions related to in-tank spray leaks for the Hazards Database and to provide more complete coverage of Tank Farm facilities. Currently, the in-tank spray leak is part of the ''Spray Leak in Structures or From Waste Transfer Lines'' accidents in Section 3.4.2.9 of the Final Safety Analysis Report (FSAR) (CHG, 2000a). The accident analysis for the ''Spray Leak in Structure or From Waste Transfer Lines'' states the following regarding the location of a possible spray leak: Inside ventilated waste storage tanks (DSTs, DCRTs, and some SSTs). Aerosols could be generated inside a storage tank during a transfer because of a leak from the portion of the transfer pipe inside the tank. The tank ventilation system could help disperse the aerosols to the atmosphere should the vent system HEPA filters fail. This Hazards Evaluation also evaluates the controls currently assigned to the spray leak in structure accident and determines the applicability of the controls to the new hazardous conditions. This comparison reviews both the analysis in the FSAR and the controls found in the Technical Safety Requirements (TSRs) (CHG, 2000h). If the new hazardous conditions do not match the analyzed accident conditions and controls, then additional analysis may be required. This document is not intended to authorize the activity or determine the adequacy of controls; it is only intended to provide information about the hazardous conditions associated with this activity. The Control decision process as defined in the AB will be used to determine the adequacy of controls and whether the proposed activity is within the AB. This hazard evaluation does not constitute an accident analysis.

GRAMS, W.H.

2000-06-13T23:59:59.000Z

7

Determination of possible damage/degradation of the Sandia National Laboratories Personal Nuclear Accident Dosimeter (PNAD).  

Science Conference Proceedings (OSTI)

This report describes the results of an inspection performed on the existing stock of SNL Personal Nuclear Accident Dosimeters (PNADs). The current stock is approximately 20 years old, and has not been examined since their initial acceptance. A small random sample of PNADs were opened (a destructive process) and the contents visually examined. Sample contents were not degraded and indicate that the existing stock of SNL PNADs is acceptable for continued use.

Potter, Charles Augustus; Ward, Dann C.

2008-05-01T23:59:59.000Z

8

Precursors to potential severe core damage accidents: 1994, a status report. Volume 22: Appendix I  

Science Conference Proceedings (OSTI)

Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 {times} 10{sup {minus}6}. These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1.

Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States); [Science Applications International Corp., Oak Ridge, TN (United States)

1995-12-01T23:59:59.000Z

9

Best Management Practice: Distribution System Audits, Leak Detection, and  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Best Management Practice: Distribution System Audits, Leak Best Management Practice: Distribution System Audits, Leak Detection, and Repair Best Management Practice: Distribution System Audits, Leak Detection, and Repair October 7, 2013 - 3:06pm Addthis A distribution system audit, leak detection, and repair programs help Federal facilities reduce water losses and make better use of limited water resources. Overview Federal facilities with large campus settings and expansive distribution systems can lose a significant amount of total water production and purchases to system leaks. Leaks in distribution systems are caused by a number of factors, including pipe corrosion, high system pressure, construction disturbances, frost damage, damaged joints, and ground shifting and settling. Regular distribution system leak detection surveys

10

Tolerating memory leaks  

Science Conference Proceedings (OSTI)

Type safety and garbage collection in managed languages eliminate memory errors such as dangling pointers, double frees, and leaks of unreachable objects. Unfortunately, a program still leaks memory if it maintains references to objects it will never ... Keywords: bug tolerance, managed languages, memory leaks

Michael D. Bond; Kathryn S. McKinley

2008-10-01T23:59:59.000Z

11

HRT LEAK DETECTION SYSTEM  

SciTech Connect

All HRT process piping and equipment is contained in a large tank and flanged connections with stainless steel ring gaskets are used where needed to permit the removal of values and items of equipment. Underwater remote maintenance is to be employed and special provisions are required for indicating and locating leaks at all mechanical joints in the process system. Each joint is monitored and a signal is given when a leak occurs. The valve operator stems are sealed with stainless steel bellows and a means of detecting a leak in the bellows has been included. (auth)

Kuster, J.E.

1956-04-20T23:59:59.000Z

12

Precursors to potential severe core damage accidents: 1992, A status report. Volume 17, Main report and Appendix A  

SciTech Connect

Twenty-seven operational events with conditional probabilities of subsequent severe core damage of 1.0 {times} 10E-06 or higher occurring at commercial light-water reactors during 1992 are considered to be precursors to potential core damage. These are described along with associated significance estimates, categorization, and subsequent analyses. The report discusses (1) the general rationale for this study, (2) the selection and documentation of events as precursors, (3) the estimation and use of conditional probabilities of subsequent severe core damage to rank precursor events, and (4) the plant models used in the analysis process.

Cox, D.F.; Cletcher, J.W.; Copinger, D.A.; Cross-Dial, A.E.; Morris, R.H.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Jansen, J.M.; Minarick, J.W. [Science Applications International Corp., Oak Ridge, TN (United States); Lau, W.; Salyer, W.D. [Reliability and Performance Associates (United States)

1993-12-01T23:59:59.000Z

13

Fixing file descriptor leaks  

E-Print Network (OSTI)

We design, implement and test a tool for eliminating file descriptor (FD) leaks in programs at run-time. Our tool monitors FD allocation and use. When the allocation of a new FD would fail because a process's entire pool ...

Dumitran, Octavian-Daniel

2007-01-01T23:59:59.000Z

14

K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations  

DOE Green Energy (OSTI)

Three bounding accidents postdated for the K West Basin integrated water treatment system are evaluated against applicable risk evaluation guidelines. The accidents are a spray leak during fuel retrieval, spray leak during backflushing, and a hydrogen explosion. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

RITTMANN, P.D.

1999-10-07T23:59:59.000Z

15

K West Basin Integrated Water Treatment System (IWTS) E-F Annular Filter Vessel Accident Calculations  

DOE Green Energy (OSTI)

Four bounding accidents postulated for the K West Basin integrated water treatment system are evaluated against applicable risk evaluation guidelines. The accidents are a spray leak during fuel retrieval, spray leak during backflushing a hydrogen explosion, and a fire breaching filter vessel and enclosure. Event trees and accident probabilities are estimated. In all cases, the unmitigated dose consequences are below the risk evaluation guidelines.

PIEPHO, M.G.

2000-01-10T23:59:59.000Z

16

APS Guideline for Accident Investigations  

NLE Websites -- All DOE Office Websites (Extended Search)

occurring in CATXSDs facilities at the APS. Definitions Accident: an unexpected event that produces personal injury, illness, or death; damage to or loss of property or...

17

SEALING SIMULATED LEAKS  

Science Conference Proceedings (OSTI)

This report details the testing equipment, procedures and results performed under Task 7.2 Sealing Simulated Leaks. In terms of our ability to seal leaks identified in the technical topical report, Analysis of Current Field Data, we were 100% successful. In regards to maintaining seal integrity after pigging operations we achieved varying degrees of success. Internal Corrosion defects proved to be the most resistant to the effects of pigging while External Corrosion proved to be the least resistant. Overall, with limitations, pressure activated sealant technology would be a viable option under the right circumstances.

Michael A. Romano

2004-09-01T23:59:59.000Z

18

Sensitive hydrogen leak detector  

DOE Patents (OSTI)

A sensitive hydrogen leak detector system using passivation of a stainless steel vacuum chamber for low hydrogen outgassing, a high compression ratio vacuum system, a getter operating at 77.5 K and a residual gas analyzer as a quantitative hydrogen sensor.

Myneni, Ganapati Rao (Yorktown, VA)

1999-01-01T23:59:59.000Z

19

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices E (Sections E.1--E.8). Volume 2, Part 3A  

SciTech Connect

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. The authors recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful.

Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

1994-06-01T23:59:59.000Z

20

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices A--D. Volume 2, Part 2  

Science Conference Proceedings (OSTI)

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report.

Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

1994-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendices F-H, Volume 2, Part 4  

Science Conference Proceedings (OSTI)

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis.

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States); Holmes, B. [AEA Technology, Dorset (United Kingdom)] [and others

1994-06-01T23:59:59.000Z

22

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5  

SciTech Connect

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis.

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States); Holmes, B. [AEA Technology, Dorset (United Kingdom)] [and others

1994-06-01T23:59:59.000Z

23

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix E (Sections E.9-E.16), Volume 2, Part 3B  

SciTech Connect

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis.

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States)] [and others

1994-06-01T23:59:59.000Z

24

Natural gas leak mapper  

DOE Patents (OSTI)

A system is described that is suitable for use in determining the location of leaks of gases having a background concentration. The system is a point-wise backscatter absorption gas measurement system that measures absorption and distance to each point of an image. The absorption measurement provides an indication of the total amount of a gas of interest, and the distance provides an estimate of the background concentration of gas. The distance is measured from the time-of-flight of laser pulse that is generated along with the absorption measurement light. The measurements are formated into an image of the presence of gas in excess of the background. Alternatively, an image of the scene is superimosed on the image of the gas to aid in locating leaks. By further modeling excess gas as a plume having a known concentration profile, the present system provides an estimate of the maximum concentration of the gas of interest.

Reichardt, Thomas A. (Livermore, CA); Luong, Amy Khai (Dublin, CA); Kulp, Thomas J. (Livermore, CA); Devdas, Sanjay (Albany, CA)

2008-05-20T23:59:59.000Z

25

Reducing Leaking Electricity  

NLE Websites -- All DOE Office Websites (Extended Search)

4 4 Reducing Leaking Electricity Figure 1. Full and standby power draws of some compact audio systems. A surprisingly large number of appliances-from computer peripherals to cable TV boxes to radios-consume electricity even after they have been switched off. Other appliances, such as cordless telephones, remote garage door openers, and battery chargers don't get switched off but draw power even when they are not performing their principal functions. The energy used while the appliance is switched off or not performing its primary purpose is called "standby consumption" or "leaking electricity." This consumption allows TVs, VCRs and garage-door openers to be ready for instant-on with a remote control, microwave ovens to display a digital

26

MASS SPECTROMETER LEAK  

DOE Patents (OSTI)

An improved valve is described for precisely regulating the flow of a sample fluid to be analyzed, such as in a mass spectrometer, where a gas sample is allowed to "leak" into an evacuated region at a very low, controlled rate. The flow regulating valve controls minute flow of gases by allowing the gas to diffuse between two mating surfaces. The structure of the valve is such as to prevent the corrosive feed gas from contacting the bellows which is employed in the operation of the valve, thus preventing deterioration of the bellows.

Shields, W.R.

1960-10-18T23:59:59.000Z

27

Leak test fitting  

DOE Patents (OSTI)

A hollow fitting for use in gas spectrometry leak testing of conduit joints is divided into two generally symmetrical halves along the axis of the conduit. A clip may quickly and easily fasten and unfasten the halves around the conduit joint under test. Each end of the fitting is sealable with a yieldable material, such as a piece of foam rubber. An orifice is provided in a wall of the fitting for the insertion or detection of helium during testing. One half of the fitting also may be employed to test joints mounted against a surface.

Pickett, Patrick T. (Kettering, OH)

1981-01-01T23:59:59.000Z

28

Leak test adapter for containers  

DOE Patents (OSTI)

An adapter is provided for facilitating the charging of containers and leak testing penetration areas. The adapter comprises an adapter body and stem which are secured to the container`s penetration areas. The container is then pressurized with a tracer gas. Manipulating the adapter stem installs a penetration plug allowing the adapter to be removed and the penetration to be leak tested with a mass spectrometer. Additionally, a method is provided for using the adapter. The present invention relates generally to leak test adapters, and more particularly to leak test adapters used with containers such as radioactive material shipping containers.

Hallett, B.H.; Hartley, M.S.

1995-12-31T23:59:59.000Z

29

Severe Accident Management Guidance Technical Basis Report  

Science Conference Proceedings (OSTI)

Guidance to aid operating crews in responding to a severe core damage accident was first developed as a response to the 1979 accident at Three Mile Island Unit 2. This guidance encompasses those actions that could be considered to arrest the progression of a core damage accident or to limit the extent of resulting releases of fission products. The original guidance was developed in a logical manner, starting with compiling the best information regarding severe-accident phenomena available at that ...

2012-10-31T23:59:59.000Z

30

Leaking Pipelines: Doctoral Student Family Formation  

E-Print Network (OSTI)

Sari M. Why the Academic Pipeline Leaks: Fewer Men thanone reason the academic pipeline leaks. 31 Blair-Loy, Mary.to leak out of the academic pipeline. The term academic

Serrano, Christyna M.

2008-01-01T23:59:59.000Z

31

Tips: Sealing Air Leaks | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Sealing Air Leaks Sealing Air Leaks Tips: Sealing Air Leaks May 16, 2013 - 5:03pm Addthis Sources of Air Leaks in Your Home. Areas that leak air into and out of your home cost you a lot of money. The areas listed in the illustration are the most common sources of air leaks. Sources of Air Leaks in Your Home. Areas that leak air into and out of your home cost you a lot of money. The areas listed in the illustration are the most common sources of air leaks. Air leaks can waste a lot of your energy dollars. One of the quickest energy-- and money-saving tasks you can do is caulk, seal, and weather strip all seams, cracks, and openings to the outside. Tips for Sealing Air Leaks Test your home for air tightness. On a windy day, carefully hold a lit incense stick or a smoke pen next to your windows, doors, electrical

32

Tips: Sealing Air Leaks | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Tips: Sealing Air Leaks Tips: Sealing Air Leaks Tips: Sealing Air Leaks May 16, 2013 - 5:03pm Addthis Sources of Air Leaks in Your Home. Areas that leak air into and out of your home cost you a lot of money. The areas listed in the illustration are the most common sources of air leaks. Sources of Air Leaks in Your Home. Areas that leak air into and out of your home cost you a lot of money. The areas listed in the illustration are the most common sources of air leaks. Air leaks can waste a lot of your energy dollars. One of the quickest energy-- and money-saving tasks you can do is caulk, seal, and weather strip all seams, cracks, and openings to the outside. Tips for Sealing Air Leaks Test your home for air tightness. On a windy day, carefully hold a lit incense stick or a smoke pen next to your windows, doors, electrical

33

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1. Volume 5: Analysis of core damage frequency from seismic events during mid-loop operations  

SciTech Connect

In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1) and the other at Sandia National Laboratories studying a boiling water reactor (Grand Gulf). Both the Brookhaven and Sandia projects have examined only accidents initiated by internal plant faults--so-called ``internal initiators.`` This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling shutdown conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Surry Unit 1. All of the many systems modeling assumptions, component non-seismic failure rates, and human error rates that were used in the internal-initiator study at Surry have been adopted here, so that the results of the two studies can be as comparable as possible. Both the Brookhaven study and this study examine only two shutdown plant operating states (POSs) during refueling outages at Surry, called POS 6 and POS 10, which represent mid-loop operation before and after refueling, respectively. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POSs 6 and 10. The results of the analysis are that the core-damage frequency of earthquake-initiated accidents during refueling outages in POS 6 and POS 10 is found to be low in absolute terms, less than 10{sup {minus}6}/year.

Budnitz, R.J. [Future Resources Associates, Inc., Berkeley, CA (United States); Davis, P.R. [PRD Consulting (United States); Ravindra, M.K.; Tong, W.H. [EQE International, Inc., Irvine, CA (United States)

1994-08-01T23:59:59.000Z

34

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 5: Analysis of core damage frequency from seismic events for plant operational state 5 during a refueling outage  

SciTech Connect

In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Sandia National Laboratories studying a boiling water reactor (Grand Gulf), and the other at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1). Both the Sandia and Brookhaven projects have examined only accidents initiated by internal plant faults---so-called ``internal initiators.`` This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling outage conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Grand Gulf. All of the many systems modeling assumptions, component non-seismic failure rates, and human effort rates that were used in the internal-initiator study at Grand Gulf have been adopted here, so that the results of the study can be as comparable as possible. Both the Sandia study and this study examine only one shutdown plant operating state (POS) at Grand Gulf, namely POS 5 representing cold shutdown during a refueling outage. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POS 5. The results of the analysis are that the core-damage frequency for earthquake-initiated accidents during refueling outages in POS 5 is found to be quite low in absolute terms, less than 10{sup {minus}7}/year.

Budnitz, R.J. [Future Resources Associates, Inc., Berkeley, CA (United States); Davis, P.R. [PRD Consulting (United States); Ravindra, M.K.; Tong, W.H. [EQE International, Inc., Irvine, CA (United States)

1994-08-01T23:59:59.000Z

35

Mitigated Transfer Line Leaks that Result in Surface Pools and Spray Leaks into Pits  

Science Conference Proceedings (OSTI)

This analysis provides radiological and toxicological consequence calculations for postulated mitigated leaks during transfers of six waste compositions. Leaks in Cleanout Boxes equipped with supplemental covers and leaks in pits are analyzed.

HEY, B.E.

1999-12-07T23:59:59.000Z

36

Precursors to potential severe core damage accidents: 1992, a status report; Volume 18: Appendices B, C, D, E, F, and G  

Science Conference Proceedings (OSTI)

This document is part of a report which documents 1992 operational events selected as accident sequence precursors. This report describes the 27 precursors identified from the 1992 licensee event reports. It also describe containment-related events; {open_quote}interesting{close_quote} events; potentially significant events that were considered impractical to analyze; copies of the licensee event reports which were cited in the cases above; and comments from the licensee and NRC in response to the preliminary reports.

NONE

1993-12-01T23:59:59.000Z

37

Common Errors in Leak Testing  

Science Conference Proceedings (OSTI)

Table 5   Leak tightness requirements for various products...1 ? 10 -4 Small Pipeline (gas) 1 ? 10 -5 Tanker (liquified natural gas) 1 ? 10 -6 Fine Storage tank (NH 3 ) 1 ? 10 -8 Heart pacemaker (gas) 1 ? 10 -11...

38

Stochastic Consequence Analysis for Waste Leaks  

SciTech Connect

This analysis evaluates the radiological consequences of potential Hanford Tank Farm waste transfer leaks. These include ex-tank leaks into structures, underneath the soil, and exposed to the atmosphere. It also includes potential misroutes, tank overflow

HEY, B.E.

2000-05-31T23:59:59.000Z

39

Candu 6 severe core damage accident consequence analysis for steam generator tube rupture scenario using MAAP4-CANDU V4.0.5A: preliminary results  

SciTech Connect

This paper describes the preliminary results of the consequence analysis for a generic AECL CANDU 6 station, when it undergoes a postulated, low probability Steam Generator multiple Tube Rupture (SGTR) severe accident with assumed unavailability of several critical plant safety systems. The Modular Accident Analysis Program for CANDU (MAAP4-CANDU) code was used for this analysis. The SGTR accident is assumed to begin with the guillotine rupture of 10 steam generator tubes in one steam generator in Primary Heat Transport System (PHTS) loop 1. For the reference case, the following systems were assumed unavailable: moderator and shield cooling, emergency core cooling, crash cool-down, and main and auxiliary feed water. Two additional cases were analyzed, one with the crash cool-down system available, and another with the crash cool-down and the auxiliary feed water systems available. The three scenarios considered in this study show that most of the initial fission product inventory would be retained within the containment by various fission product retention mechanisms. For the case where the crash cool-down system was credited but the auxiliary feed water systems were not credited, the total mass of volatile fission products released to the environment including stable and radioactive isotopes was about four times more than in the reference case, because fission products could be released directly from the PHTS to the environment through the Main Steam Safety Valves (MSSVs), bypassing the containment. For the case where the crash cool-down and auxiliary feed water systems were credited, the volatile fission product release to the environment was insignificant, because the fission product release was substantially mitigated by scrubbing in the water pool in the secondary side of the steam generator (SG). (authors)

Petoukhov, S.M.; Awadh, B.; Mathew, P.M. [Chalk River Laboratories, Atomic Energy of Canada Limited, Chalk River, Ontario, K0J 1J0 (Canada)

2006-07-01T23:59:59.000Z

40

Nuclear Reactor Severe Accident Experiments  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Leaking electricity in domestic appliances  

Science Conference Proceedings (OSTI)

Many types of home electronic equipment draw electric power when switched off or not performing their principal functions. Standby power use (or ''leaking electricity'') for most appliances ranges from 1 - 20 watts. Even though standby use of each device is small, the combined standby power use of all appliances in a home can easily exceed 50 watts. Leaking electricity is already responsible for 5 to 10 percent of residential electricity use in the United States and over 10 percent in Japan. An increasing number of white goods also have standby power requirements. There is a growing international effort to limit standby power to around one watt per device. New and existing technologies are available to meet this target at little or no extra cost.

Meier, Alan; Rosen, Karen

1999-05-01T23:59:59.000Z

42

Long-life leak standard assembly  

DOE Patents (OSTI)

The present invention is directed to a portable leak standard assembly which is capable of providing a stream of high-purity reference gas at a virtually constant flow rate over an extensive period of time. The leak assembly comprises a high pressure reservoir coupled to a metal leak valve through a valve-controlled conduit. A reproducible leak valve useful in this assembly is provided by a metal tube crimped with a selected pressure loading for forming an orifice in the tube with this orifice being of a sufficient size to provide the selected flow rate. The leak valve assembly is formed of metal so that it can be "baked-out" in a vacuum furnace to rid the reservoir and attendent components of volatile impurities which reduce the efficiency of the leak standard.

Basford, James A. (Oak Ridge, TN); Mathis, John E. (Oak Ridge, TN); Wright, Harlan C. (Oak Ridge, TN)

1982-01-01T23:59:59.000Z

43

High sensitivity leak detection method and apparatus  

DOE Patents (OSTI)

An improved leak detection method is provided that utilizes the cyclic adsorption and desorption of accumulated helium on a non-porous metallic surface. The method provides reliable leak detection at superfluid helium temperatures. The zero drift that is associated with residual gas analyzers in common leak detectors is virtually eliminated by utilizing a time integration technique. The sensitivity of the apparatus of this disclosure is capable of detecting leaks as small as 1 [times] 10[sup [minus]18] atm cc sec[sup [minus]1]. 2 figs.

Myneni, G.R.

1994-09-06T23:59:59.000Z

44

High sensitivity leak detection method and apparatus  

DOE Patents (OSTI)

An improved leak detection method is provided that utilizes the cyclic adsorption and desorption of accumulated helium on a non-porous metallic surface. The method provides reliable leak detection at superfluid helium temperatures. The zero drift that is associated with residual gas analyzers in common leak detectors is virtually eliminated by utilizing a time integration technique. The sensitivity of the apparatus of this disclosure is capable of detecting leaks as small as 1.times.10.sup.-18 atm cc sec.sup.-1.

Myneni, Ganapatic R. (Grafton, VA)

1994-01-01T23:59:59.000Z

45

Nuclear Reactor Accidents  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

46

Leak checker data logging system  

DOE Patents (OSTI)

A portable, high speed, computer-based data logging system for field testing systems or components located some distance apart employs a plurality of spaced mass spectrometers and is particularly adapted for monitoring the vacuum integrity of a long string of a superconducting magnets such as used in high energy particle accelerators. The system provides precise tracking of a gas such as helium through the magnet string when the helium is released into the vacuum by monitoring the spaced mass spectrometers allowing for control, display and storage of various parameters involved with leak detection and localization. A system user can observe the flow of helium through the magnet string on a real-time basis hour the exact moment of opening of the helium input valve. Graph reading can be normalized to compensate for magnet sections that deplete vacuum faster than other sections between testing to permit repetitive testing of vacuum integrity in reduced time.

Payne, J.J.; Gannon, J.C.

1994-12-31T23:59:59.000Z

47

Leak checker data logging system  

DOE Patents (OSTI)

A portable, high speed, computer-based data logging system for field testing systems or components located some distance apart employs a plurality of spaced mass spectrometers and is particularly adapted for monitoring the vacuum integrity of a long string of a superconducting magnets such as used in high energy particle accelerators. The system provides precise tracking of a gas such as helium through the magnet string when the helium is released into the vacuum by monitoring the spaced mass spectrometers allowing for control, display and storage of various parameters involved with leak detection and localization. A system user can observe the flow of helium through the magnet string on a real-time basis hour the exact moment of opening of the helium input valve. Graph reading can be normalized to compensate for magnet sections that deplete vacuum faster than other sections between testing to permit repetitive testing of vacuum integrity in reduced time.

Gannon, Jeffrey C. (Arlington, TX); Payne, John J. (Waterman, IL)

1996-01-01T23:59:59.000Z

48

Leak checker data logging system  

DOE Patents (OSTI)

A portable, high speed, computer-based data logging system for field testing systems or components located some distance apart employs a plurality of spaced mass spectrometers and is particularly adapted for monitoring the vacuum integrity of a long string of a superconducting magnets such as used in high energy particle accelerators. The system provides precise tracking of a gas such as helium through the magnet string when the helium is released into the vacuum by monitoring the spaced mass spectrometers allowing for control, display and storage of various parameters involved with leak detection and localization. A system user can observe the flow of helium through the magnet string on a real-time basis hour the exact moment of opening of the helium input valve. Graph reading can be normalized to compensate for magnet sections that deplete vacuum faster than other sections between testing to permit repetitive testing of vacuum integrity in reduced time. 18 figs.

Gannon, J.C.; Payne, J.J.

1996-09-03T23:59:59.000Z

49

One watt initiative: A global effort to reduce leaking electricity  

E-Print Network (OSTI)

National Laboratory - Leaking Electricity Web Site http://Effort to Reduce Leaking Electricity Alan MEIER* & Benotfraction of total electricity use. Several initiatives to

Meier, Alan K.; LeBot, Benoit

1999-01-01T23:59:59.000Z

50

Modular, High-Volume Fuel Cell Leak-Test Suite and Process  

DOE Green Energy (OSTI)

Fuel cell stacks are typically hand-assembled and tested. As a result the manufacturing process is labor-intensive and time-consuming. The fluid leakage in fuel cell stacks may reduce fuel cell performance, damage fuel cell stack, or even cause fire and become a safety hazard. Leak check is a critical step in the fuel cell stack manufacturing. The fuel cell industry is in need of fuel cell leak-test processes and equipment that is automatic, robust, and high throughput. The equipment should reduce fuel cell manufacturing cost.

Ru Chen; Ian Kaye

2012-03-12T23:59:59.000Z

51

DIY BASICS CHECKLIST DRIPS AND LEAKS  

E-Print Network (OSTI)

DIY BASICS CHECKLIST DRIPS AND LEAKS Watercancauseseriousdamage- oftenunseen. Drillbits. Tapemeasure. Spiritlevel. Start off small. Collect a basic tool kit. There's plenty of DIY info'tdrillintomortarbetweenbricks. #12;DIY BASICS CHECKLIST Location Twopeoplemakethisamuch easierjob. Cutasheetofpapertothesize

Peters, Richard

52

Cork: dynamic memory leak detection for garbage-collected languages  

Science Conference Proceedings (OSTI)

A memory leak in a garbage-collected program occurs when the program inadvertently maintains references to objects that it no longer needs. Memory leaks cause systematic heap growth, degrading performance and resulting in program crashes after ... Keywords: dynamic, garbage collection, memory leak detection, memory leaks, runtime analysis

Maria Jump; Kathryn S. McKinley

2007-01-01T23:59:59.000Z

53

New findings on leak resistance of API 8-Round connectors  

Science Conference Proceedings (OSTI)

In response to high interest concerning leak resistance in API 8-Round connectors, the API funded projects that have identified and assessed parameters affecting leak. Among these parameters are make-up, diameter, grade, and combined loads. Additional turns during make-up was found to increase leak resistance. Investigations concerning diameter and grade identified larger diameter and higher grade connectors as most susceptible to low leak pressures when compared to pipe body ratings. Finally, combined loads were found to be crucial to leak. Tension lowers the leak resistance of 8-Round connectors in a manner that renders hydrotesting insufficient for defining leak in some service conditions.

Schwind, B.E.; Wooley, G.R.

1986-01-01T23:59:59.000Z

54

SINGLE-SHELL TANKS LEAK INTEGRITY ELEMENTS/SX FARM LEAK CAUSES AND LOCATIONS - 12127  

SciTech Connect

Washington River Protection Solutions, LLC (WRPS) developed an enhanced single-shell tank (SST) integrity project in 2009. An expert panel on SST integrity was created to provide recommendations supporting the development of the project. One primary recommendation was to expand the leak assessment reports (substitute report or LD-1) to include leak causes and locations. The recommendation has been included in the M-045-9IF Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) as one of four targets relating to SST leak integrity. The 241-SX Farm (SX Farm) tanks with leak losses were addressed on an individual tank basis as part of LD-1. Currently, 8 out of 23 SSTs that have been reported to having a liner leak are located in SX Farm. This percentage was the highest compared to other tank farms which is why SX Farm was analyzed first. The SX Farm is comprised of fifteen SSTs built 1953-1954. The tanks are arranged in rows of three tanks each, forming a cascade. Each of the SX Farm tanks has a nominal I-million-gal storage capacity. Of the fifteen tanks in SX Farm, an assessment reported leak losses for the following tanks: 241-SX-107, 241-SX-108, 241-SX-109, 241-SX-111, 241-SX-112, 241-SX-113, 241-SX-114 and 241-SX-115. The method used to identify leak location consisted of reviewing in-tank and ex-tank leak detection information. This provided the basic data identifying where and when the first leaks were detected. In-tank leak detection consisted of liquid level measurement that can be augmented with photographs which can provide an indication of the vertical leak location on the sidewall. Ex-tank leak detection for the leaking tanks consisted of soil radiation data from laterals and drywells near the tank. The in-tank and ex-tank leak detection can provide an indication of the possible leak location radially around and under the tank. Potential leak causes were determined using in-tank and ex-tank information that is not directly related to leak detection. In-tank parameters can include temperature of the supernatant and sludge, types of waste, and chemical determination by either transfer or sample analysis. Ex-tank information can be assembled from many sources including design media, construction conditions, technical specifications, and other sources. Five conditions may have contributed to SX Farm tank liner failure including: tank design, thermal shock, chemistry-corrosion, liner behavior (bulging), and construction temperature. Tank design did not apparently change from tank to tank for the SX Farm tanks; however, there could be many unknown variables present in the quality of materials and quality of construction. Several significant SX Farm tank design changes occurred from previous successful tank farm designs. Tank construction occurred in winter under cold conditions which could have affected the ductile to brittle transition temperature of the tanks. The SX Farm tanks received high temperature boiling waste from REDOX which challenged the tank design with rapid heat up and high temperatures. All eight of the leaking SX Farm tanks had relatively high rate of temperature rise. Supernatant removal with subsequent nitrate leaching was conducted in all but three of the eight leaking tanks prior to leaks being detected. It is possible that no one characteristic of the SX Farm tanks could in isolation from the others have resulted in failure. However, the application of so many stressors - heat up rate, high temperature, loss of corrosion protection, and tank design - working jointly or serially resulted in their failure. Thermal shock coupled with the tank design, construction conditions, and nitrate leaching seem to be the overriding factors that can lead to tank liner failure. The distinction between leaking and sound SX Farm tanks seems to center on the waste types, thermal conditions, and nitrate leaching.

VENETZ TJ; WASHENFELDER D; JOHNSON J; GIRARDOT C

2012-01-25T23:59:59.000Z

55

Accident Investigation Handbook  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Improvement (HPI). The recommended techniques apply equally well to DOE Federal-led accident investigations conducted under DOE Order (O) 225.1B, Accident Investigations,...

56

Location of Leaks in Pressure Testable Direct Burial Steam Distribution Conduits  

E-Print Network (OSTI)

Central steam is commonly distributed through direct burial lines protected by an outer conduit. These underground conduit systems are subject to electrolytic corrosion. Failure of the outer casing permits water intrusion and damage to insulation, resulting in increased thermal energy losses and eventual damage to the steam line. Breaches in the outer conduit are difficult to locate, and damage to the steam line may progress until the entire line requires replacement. Thermal energy losses are high if groundwater infiltrates the conduit and excavation to replace the steam line is extremely expensive. Locating leaks in steam line conduit is a two step procedure. The first step is to regularly pressure test sections of conduit to determine whether a breach has occurred. Pressure testing should be performed on a regular basis to minimize thermal losses and damage from groundwater intrusion. If pressure testing reveals that the conduit is leaking, the Navy has developed a procedure and equipment to determine where the breach occurred. The breach can be detected using sulfur hexafluoride (SF6) tracer gas injected into the conduit. After injection, maintenance personnel walk the path of the steam line with an SF6 detector that precisely locates the leak. Then, only the necessary conduit sections are excavated for repair. We have successfully used this system at several locations, and in a variety of soil conditions. Tracer gas leak testing provides an effective and inexpensive method to evaluate underground conduit systems. Performed on a regular basis, it is a useful preventive maintenance tool to minimize energy loss and utility system damage. Test results also provide valuable input to the decision to repair or replace underground steam lines. This equipment and procedure may be used on other utility system distribution components, such as compressed air and direct burial steam lines.

Sittel, M. G.; Messock, R. K.

1993-03-01T23:59:59.000Z

57

Accident progression event tree analysis for postulated severe accidents at N Reactor  

SciTech Connect

A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied.

Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. (Sandia National Labs., Albuquerque, NM (USA)); Medford, G.T. (Science Applications International Corp., Albuquerque, NM (USA))

1990-06-01T23:59:59.000Z

58

Detecting Air Leaks | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Detecting Air Leaks Detecting Air Leaks Detecting Air Leaks September 27, 2012 - 6:39pm Addthis For a thorough and accurate measurement of air leakage in your home, hire a qualified technician to conduct an energy assessment, particularly a blower door test. For a thorough and accurate measurement of air leakage in your home, hire a qualified technician to conduct an energy assessment, particularly a blower door test. You may already know where some air leakage occurs in your home, such as an under-the-door draft, but you'll need to find the less obvious gaps to properly air seal your home. For a thorough and accurate measurement of air leakage in your home, hire a qualified technician to conduct an energy assessment, particularly a blower door test. A blower door test, which depressurizes a home, can

59

EPR Severe Accident Threats and Mitigation  

SciTech Connect

Despite the extremely low EPR core melt frequency, an improved defence-in-depth approach is applied in order to comply with the EPR safety target: no stringent countermeasures should be necessary outside the immediate plant vicinity like evacuation, relocation or food control other than the first harvest in case of a severe accident. Design provisions eliminate energetic events and maintain the containment integrity and leak-tightness during the entire course of the accident. Based on scenarios that cover a broad range of physical phenomena and which provide a sound envelope of boundary conditions associated with each containment challenge, a selection of representative loads has been done, for which mitigation measures have to cope with. This paper presents the main critical threats and the approach used to mitigate those threats. (authors)

Azarian, G. [Framatome ANP SAS, Tour Areva, Place de la Coupole 92084 Paris la Defense (France); Kursawe, H.M.; Nie, M.; Fischer, M.; Eyink, J. [Framatome ANP GmbH, Freyeslebenstrasse, 1, D-91058 Erlangen (Germany); Stoudt, R.H. [Framatome ANP Inc. - 3315 Old Forest Rd, Lynchburgh, VA 24501 (United States)

2004-07-01T23:59:59.000Z

60

B Plant ion exchange feed line leak  

SciTech Connect

>One of the objectives of the Waste Management Program is to separate the long-lived heat emitter /aup 137/Cs from the bulk of the high-level Iiquid wastes. This separation is accomplished by the ion exchange process in the 221-B Building. Interim storage of the cesium is in solution as a nitrate. The feed for the B Plant cesinm ion exchange process is pumped from the lag storage tank, 105-C, through a pipeline and several diversion boxes to the 221-B Building. On December 19, 1969, a leak was discovered near the 241-C-152 diversion box in the section of this line, V-122, from the 105-C tank. Although the leak represented a loss of feed for the processing of /sup 137/Cs, more important was the consequence of environmental contmination to the soil from the line leak. For this reason, an investigation was made to estblish the extent of the radioactivity spread. The results of a well drilling operation undertaken to define the boundary and to estimate the extent of the leak are summarized. (CR)

Tanaka, K.H.

1971-01-25T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Evaluation of Accident Frequencies at the Canister Storage Bldg (CSB)  

DOE Green Energy (OSTI)

By using simple frequency calculations and fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The following are the design basis accidents: Mechanical damage of MCO; Gaseous release from the MCO; MCO internal hydrogen deflagration; MCO external hydrogen deflagration; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

POWERS, T.B.

2000-03-20T23:59:59.000Z

62

Savings Project: How to Seal Air Leaks with Caulk | Department...  

NLE Websites -- All DOE Office Websites (Extended Search)

How to Seal Air Leaks with Caulk Addthis Project Level Easy Energy Savings 5 - 10% Time to Complete 1 - 2 Hours Overall Cost 3 - 30 Sealing air leaks around windows and...

63

Modeling leaks from liquid hydrogen storage systems.  

DOE Green Energy (OSTI)

This report documents a series of models for describing intended and unintended discharges from liquid hydrogen storage systems. Typically these systems store hydrogen in the saturated state at approximately five to ten atmospheres. Some of models discussed here are equilibrium-based models that make use of the NIST thermodynamic models to specify the states of multiphase hydrogen and air-hydrogen mixtures. Two types of discharges are considered: slow leaks where hydrogen enters the ambient at atmospheric pressure and fast leaks where the hydrogen flow is usually choked and expands into the ambient through an underexpanded jet. In order to avoid the complexities of supersonic flow, a single Mach disk model is proposed for fast leaks that are choked. The velocity and state of hydrogen downstream of the Mach disk leads to a more tractable subsonic boundary condition. However, the hydrogen temperature exiting all leaks (fast or slow, from saturated liquid or saturated vapor) is approximately 20.4 K. At these temperatures, any entrained air would likely condense or even freeze leading to an air-hydrogen mixture that cannot be characterized by the REFPROP subroutines. For this reason a plug flow entrainment model is proposed to treat a short zone of initial entrainment and heating. The model predicts the quantity of entrained air required to bring the air-hydrogen mixture to a temperature of approximately 65 K at one atmosphere. At this temperature the mixture can be treated as a mixture of ideal gases and is much more amenable to modeling with Gaussian entrainment models and CFD codes. A Gaussian entrainment model is formulated to predict the trajectory and properties of a cold hydrogen jet leaking into ambient air. The model shows that similarity between two jets depends on the densimetric Froude number, density ratio and initial hydrogen concentration.

Winters, William Stanley, Jr.

2009-01-01T23:59:59.000Z

64

Low level waste shipment accident lessons learned  

SciTech Connect

On October 1, 1994 a shipment of low-level waste from the Fernald Environmental Management Project, Fernald, Ohio, was involved in an accident near Rolla, Missouri. The accident did not result in the release of any radioactive material. The accident did generate important lessons learned primarily in the areas of driver and emergency response communications. The shipment was comprised of an International Standards Organization (ISO) container on a standard flatbed trailer. The accident caused the low-level waste package to separate from the trailer and come to rest on its top in the median. The impact of the container with the pavement and median inflicted relatively minor damage to the container. The damage was not substantial enough to cause failure of container integrity. The success of the package is attributable to the container design and the packaging procedures used at the Fernald Environmental Management Project for low-level waste shipments. Although the container survived the initial wreck, is was nearly breached when the first responders attempted to open the ISO container. Even though the container was clearly marked and the shipment documentation was technically correct, this information did not identify that the ISO container was the primary containment for the waste. The lessons learned from this accident have DOE complex wide applicability. This paper is intended to describe the accident, subsequent emergency response operations, and the lessons learned from this incident.

Rast, D.M.; Rowe, J.G.; Reichel, C.W.

1995-02-01T23:59:59.000Z

65

Novel NIST Connector Uses Magnets for Leak-Free ...  

Science Conference Proceedings (OSTI)

Novel NIST Connector Uses Magnets for Leak-Free Microfluidic Devices. For Immediate Release: November 17, 2009. ...

2012-10-15T23:59:59.000Z

66

Preliminary analysis of tank 241-C-106 dryout due to large postulated leak and vaporization  

SciTech Connect

This analysis assumes that there is a hypothetical large leak at the bottom of Tank 241-C-106 which initiates the dryout of the tank. The time required for a tank to dryout after a leak is of interest for safety reasons. As a tank dries out, its temperature is expected to increase which could affect the structural integrity of the concrete tank dome. Hence, it is of interest to know how fast and how high the temperature in a leaky tank increases, so that mitigation procedures can be planned and implemented in a timely manner. This analysis is focused on tank 241-C-106, which is known to be high thermal tank. The objective of the study was to determine how long it would take for tank 241-C-106 to reach 350 degrees Fahrenheit (about 177 degrees Centigrade) after a postulated large leak develops at the bottom center of the tank. The temperature of 350 degrees Fahrenheit is the minimum temperature that can cause structural damage to concrete (ACI 1992). The postulated leak at the bottom of the tank and the resulting dryout of the sludge in the tank make this analysis different from previous thermal analyses of the C-106 tank and other tanks, especially the double-shell tanks which are mostly liquid.

Piepho, M.G.

1994-12-01T23:59:59.000Z

67

Slowing leaking electricity to a trickle  

SciTech Connect

Electronics play an increasingly pervasive role in home appliances and office equipment. This is generally a good thing because the electronics help provide new features and amenities. Electronic controls can also reduce energy use by providing the services only when consumers actually need them. On the other hand, these electronic features often continue to consume energy even while switched off or not performing their principal service. The technical term for this phenomenon is ''standby power consumption'' but it has acquired several common names, including ''leaking electricity,'' ''waiting electricity,'' ''free-running power,'' ''off-mode power,'' and ''phantom loads.'' The leaking electricity found in our televisions, VCRs, garage door openers, cordless phones and many other appliances has a surprisingly large impact on the global environment.

Meier, Alan

1998-07-01T23:59:59.000Z

68

Method for mapping a natural gas leak  

DOE Patents (OSTI)

A system is described that is suitable for use in determining the location of leaks of gases having a background concentration. The system is a point-wise backscatter absorption gas measurement system that measures absorption and distance to each point of an image. The absorption measurement provides an indication of the total amount of a gas of interest, and the distance provides an estimate of the background concentration of gas. The distance is measured from the time-of-flight of laser pulse that is generated along with the absorption measurement light. The measurements are formatted into an image of the presence of gas in excess of the background. Alternatively, an image of the scene is superimposed on the image of the gas to aid in locating leaks. By further modeling excess gas as a plume having a known concentration profile, the present system provides an estimate of the maximum concentration of the gas of interest.

Reichardt, Thomas A. (Livermore, CA); Luong, Amy Khai (Dublin, CA); Kulp, Thomas J. (Livermore, CA); Devdas, Sanjay (Albany, CA)

2009-02-03T23:59:59.000Z

69

Planning and care mark repair of 14-year old leak in Kuwait Oil Co. LPG tank 95  

SciTech Connect

This paper points out that the leak, which had been present for such a long time, completely saturated the perlite insulation with hydrocarbons, thus rendering the entire operation of inspection, repair, and maintenance of the inner tank a hazardous operation. It emphasizes the safety aspects, which were complicated by the saturated perlite as well as by the fact that the tank is situated in the middle of the LPG storage area with LPG tanks on either side. Tank design, making preparations, inspection, and repair are discussed. The fact that the leaking flanges were originally installed damaged, indicated the future need of tighter company quality control of all contractors work.

Shtayieh, S.

1983-01-10T23:59:59.000Z

70

Probabilistic pipe fracture evaluations for leak-rate-detection applications  

SciTech Connect

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crack size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.

Rahman, S.; Ghadiali, N.; Paul, D.; Wilkowski, G. [Battelle, Columbus, OH (United States)

1995-04-01T23:59:59.000Z

71

Distribution System Audits, Leak Detection, and Repair: Kirtland Air Force Base Leak Detection and Repair Program  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Best Management Practice Best Management Practice Case Study #3 Distribution System Audits, Leak Detection, and Repair Kirtland Air Force Base - Leak Detection and Repair Program Overview Kirtland Air Force Base (AFB) performed an award winning leak detection and repair program in 2006. The results of the project are saving Kirtland AFB 179 million gallons each year, which is over 16% of the total water use at the base. Kirtland AFB is located on 52,000 acres, southeast and adjacent to Albuquerque, New Mexico. The area is a high altitude desert, only receiving about 8 inches of rain each year. Kirtland AFB draws water from an under- ground aquifer via seven production wells through- out the base. The base also has access to water from the City of Albuquerque. The underground water

72

Distribution System Audits, Leak Detection, and Repair: Kirtland Air Force Base Leak Detection and Repair Program  

NLE Websites -- All DOE Office Websites (Extended Search)

Best Management Practice Best Management Practice Case Study #3 Distribution System Audits, Leak Detection, and Repair Kirtland Air Force Base - Leak Detection and Repair Program Overview Kirtland Air Force Base (AFB) performed an award winning leak detection and repair program in 2006. The results of the project are saving Kirtland AFB 179 million gallons each year, which is over 16% of the total water use at the base. Kirtland AFB is located on 52,000 acres, southeast and adjacent to Albuquerque, New Mexico. The area is a high altitude desert, only receiving about 8 inches of rain each year. Kirtland AFB draws water from an under- ground aquifer via seven production wells through- out the base. The base also has access to water from the City of Albuquerque. The underground water

73

Mitigation of Nuclear Fuel Pool Leaks  

Science Conference Proceedings (OSTI)

The used or spent fuel from nuclear reactors is stored in spent fuel pools, which require canals for fuel transfer activities. These pools--3540 feet or more in depth--are lined with stainless steel ranging in thickness from ~.19 in~.38 in (~4.8 mm~9.5 mm). The liners are anchored to the walls and slab via welds that can leak or crack. lectricit de France (EDF) has developed tools to check suspect areas of the liner seam welds for cracking or leakage. This report ...

2013-08-29T23:59:59.000Z

74

One-Piece Leak-Proof Battery  

SciTech Connect

The casing of a leak-proof one-piece battery is made of a material comprising a mixture of at least a matrix based on polypropylene and an alloy of a polyamide and a polypropylene. The ratio of the matrix to the alloy is in the range 0.5 to 6 by weight. The alloy forms elongate arborescent inclusions in the matrix such that, on average, the largest dimension of a segment of the arborescence is at least twenty times the smallest dimension of the segment.

Verhoog, Roelof (Bordeaux, FR)

1999-03-23T23:59:59.000Z

75

Low heat-leak cryogenic envelope  

DOE Patents (OSTI)

A plurality of cryogenic envelope sections are joined together to form a power transmission line. Each of the sections is comprised of inner and outer tubes having multilayer metalized plastic spirally wrapped within a vacuum chamber formed between the inner and outer tubes. A refrigeration tube traverses the vacuum chamber, but exits one section and enters another through thermal standoffs for reducing heat-leak from the outer tube to the refrigeration tube. The refrigeration tube passes through a spirally wrapped shield within each section's vacuum chamber in a manner so that the refrigeration tube is in close thermal contact with the shield, but is nevertheless slideable with respect thereto.

DeHaan, James R. (Boulder, CO)

1976-10-19T23:59:59.000Z

76

Long-life leak standard assembly. [Patent application  

DOE Patents (OSTI)

The present invention is directed to a portable leak standard assembly which is capable of providing a stream of high-purity reference gas at a virtually constant flow rate over an extensive period of time. The leak assembly comprises a high pressure reservoir coupled to a metal leak valve through a valve-controlled conduit. A reproducible leak valve useful in this assembly is provided by a metal tube crimped with a selected pressure loading for forming an orifice in the tube with this orifice being of a sufficient size to provide the selected flow rate. The leak valve assembly is formed of metal so that it can be baked-out in a vacuum furnace to rid the reservoir and attendent components of volatile impurities which reduce the efficiency of the leak standard.

Basford, J.A.; Mathis, J.E.; Wright, H.C.

1980-11-12T23:59:59.000Z

77

Significant factors in rail freight accidents: A statistical analysis of predictive and severity indices in the FRA accident/incident data base  

Science Conference Proceedings (OSTI)

The Federal Railroad Association maintains a file of carrier-reported accidents and incidents that meet threshold criteria for damage cost and/or casualties. Using a five year period from this data base, an investigation was conducted into the relationship between quantifiable risk factors and accident frequency and severity. Specific objectives were to identify key variables in accidents, formulate a model to predict future accidents, and assess the relative importance of these variables from the perspective of routing and shipping decision making. The temporal factors YEAR and MONTH were found to be significant predictors of risk; accident severity was greatest for accidents caused by track and roadbed defects. Train speed was an indicator of accident severity; track class and training tonnage were inversely proportional to accident severity. Investigation of the data base is continuing, with a final report expected by late summer. 15 refs., 1 fig., 10 tabs.

Lee, Tze-San; Saricks, C.L.

1991-01-01T23:59:59.000Z

78

Savings Project: How to Seal Air Leaks with Caulk | Department...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Electricity Rates Savings Project: How to Seal Air Leaks with Caulk Tips: Windows Household Heating Systems: Although several different types of fuels are available to heat...

79

Hydrogen leak detection - low cost distributed gas sensors  

NLE Websites -- All DOE Office Websites (Extended Search)

leak detection that can be economically satisfied using our technology. * Due to limited refinery capacity, downtime in the oil and gas refining industry has become of critical...

80

Distributed Optical Sensor for CO2 Leak Detection  

NLE Websites -- All DOE Office Websites (Extended Search)

on the technology "Distributed Optical Sensor for CO 2 Leak Detection," for which a Patent Application has been filed. This technology is available for licensing andor further...

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Nuclear criticality accidents  

SciTech Connect

Criticality occurs when a sufficient quantity of fissionable material is accumulated, and it results in the liberation of nuclear energy. All process accidents have involved plutonium or highly enriched uranium, as have most of the critical experiment accidents. Slightly enriched uranium systems require much larger quantities of material to achieve criticality. An appreciation of criticality accidents should be based on an understanding of factors that influence criticality, which are discussed in this article. 11 references.

Smith, D.R. (Los Alamos National Laboratory, New Mexico (Unites States))

1991-10-01T23:59:59.000Z

82

Severe Accident Studies | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Severe Accident Studies Severe Accident Studies Powerpoint discussing studies and conclusions on transportation accidents and safety. Severe Accident Studies More Documents &...

83

Fukushima Daiichi Accident -- Technical Causal Factor Analysis  

Science Conference Proceedings (OSTI)

On March 11, 2011, the Fukushima Daiichi nuclear power plant experienced a seismic event and subsequent tsunami. The accident and the ensuing mitigation and recovery activities occurred over several days, involved a number of incidents, and might provide several opportunities for lessons learned. The objective of this report is to determine the fundamental causative factors for the loss of critical systems at the Fukushima Daiichi reactors that resulted in core damage and subsequent radioactive release. ...

2012-03-27T23:59:59.000Z

84

A new blowdown compensation scheme for boiler leak detection  

E-Print Network (OSTI)

A new blowdown compensation scheme for boiler leak detection A. M. Pertew ,1 X. Sun ,1 R. Kent considers the blowdown effect in industrial boiler operation. This adds to the efficiency of recent advances in identification-based leak detection techniques of boiler steam- water systems. Keywords: Industrial Boilers, Tube

Marquez, Horacio J.

85

Commercial Grade Item (CGI) Dedication for Leak Detection Relays  

Science Conference Proceedings (OSTI)

This Test Plan provides a test method to dedicate the leak detection relays used on the new Pumping and Instrumentation Control (PIC) skids. The new skids are fabricated on-site. The leak detection system is a safety class system per the Authorization Basis.

KOCH, M.R.

1999-10-26T23:59:59.000Z

86

Commercial Grade Item (CGI) Dedication for Leak Detection Relays  

Science Conference Proceedings (OSTI)

This Test Plan provides a test method to dedicate the leak detection relays used on the new Pumping Instrumentation and Control (PIC) skids. The new skids are fabricated on-site. The leak detection system is a safety class system per the Authorization Basis.

JOHNS, B.R.; KOCH, M.R.

2000-01-28T23:59:59.000Z

87

Commercial Grade Item (CGI) Dedication for Leak Detection Relays  

SciTech Connect

This Test Plan provides a test method to dedicate the leak detection relays used on the new Pumping and Instrumentation Control (PIC) skids. The new skids are fabricated on-site. The leak detection system is a safety class system per the Authorization Basis.

KOCH, M.R.; JOHNS, B.R.

1999-12-21T23:59:59.000Z

88

Commercial Grade Item (CGI) Dedication for Leak Detection Relays  

Science Conference Proceedings (OSTI)

This Test Plan provides a test method to dedicate the leak detection relays used on the new Pumping Instrumentation and Control (PIC) skids. The new skids are fabricated on-site. The leak detection system is a safety class system per the Authorization Basis.

KOCH, M.R.

2000-02-28T23:59:59.000Z

89

Commercial Grade Item (CGI) Dedication for Leak Detection Relays  

SciTech Connect

This Test Plan provides a test method to dedicate the leak detection relays used on the new Pumping and Instrumentation Control (PIC) skids. The new skids are fabricated on-site. The leak detection system is a safety class system per the Authorization Basis.

KOCH, M.R.

1999-08-11T23:59:59.000Z

90

Heat exchanger with leak detecting double wall tubes  

DOE Patents (OSTI)

A straight shell and tube heat exchanger utilizing double wall tubes and three tubesheets to ensure separation of the primary and secondary fluid and reliable leak detection of a leak in either the primary or the secondary fluids to further ensure that there is no mixing of the two fluids.

Bieberbach, George (Tampa, FL); Bongaards, Donald J. (Seminole, FL); Lohmeier, Alfred (Tampa, FL); Duke, James M. (St. Petersburg, all of, FL)

1981-01-01T23:59:59.000Z

91

Gas Leak from Vinyl Taped Stainless Steel Dressing Jars  

DOE Green Energy (OSTI)

The leak rates of nitrogen gas from stainless steel dressing jars taped with 2 inch vinyl tape were measured. These results were used to calculate hydrogen leak rates from the same jars. The calculations show that the maximum concentration of hydrogen buildup in this type of container configuration will beat least 3 orders of magnitude below the lower explosion limit for hydrogen in air.

Tim Hayes

1999-03-01T23:59:59.000Z

92

Mineral formation during simulated leaks of Hanford waste tanks  

E-Print Network (OSTI)

Mineral formation during simulated leaks of Hanford waste tanks Youjun Deng a , James B. Harsh a at the US DOE Hanford Site, Washington, caus- ing mineral dissolution and re-precipitation upon contact mimicking tank leak conditions at the US DOE Hanford Site. In batch experiments, Si-rich solutions

Flury, Markus

93

Detecting leaks to reduce energy costs  

SciTech Connect

This article describes how analyzing boilerhouse data in its manufacturing plants and applying algorithmic techniques is helping an automobile manufacturer run its utility operations more efficiently. Ford Motor Co., based in Dearborn, Michigan, is realizing significant energy savings, reducing capital expenditures, and minimizing wastewater disposal costs by diagnosing and quantifying leaks in its compressed air, steam/condensate, and process water systems by applying algorithms developed by Cleveland-based CEC Consultants Inc. These algorithms make use of readily available--and often already installed--instruments, such as vortex shedding meters, chart recorders, and data loggers, to compare how much utility use is needed for assembly and manufacturing equipment with how much is being generated.

Valenti, M.

1995-07-01T23:59:59.000Z

94

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Analysis of core damage frequency from internal events for plant operational state 5 during a refueling outage. Internal events appendices K to M  

Science Conference Proceedings (OSTI)

This report provides supporting documentation for various tasks associated with the performance of the probabilistic risk assessment for Plant Operational State 5 (approximately Cold Shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage at Grand Gulf, Unit 1 as documented in Volume 2, Part 1 of NUREG/CR-6143. The report contains the following appendices: K - HEP Locator Files; L - Supporting Information for the Plant Damage State Analysis; M - Summary of Results from the Coarse Screening Analysis - Phase 1A.

Forester, J.; Yakle, J.; Walsh, B. [Science Applications International Corp., Albuquerque, NM (United States); Darby, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States); Whitehead, D.; Staple, B.; Brown, T. [Sandia National Labs., Albuquerque, NM (United States)

1994-07-01T23:59:59.000Z

95

Proceedings of the seminar on leak before break in reactor piping and vessels  

SciTech Connect

The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Faidy, C. [ed.] [Electricite de France, Villeurbanne (France); Gilles, P. [ed.] [Framatome, Paris (France)

1997-04-01T23:59:59.000Z

96

Severe Accident Studies  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Severe Accident Studies Severe Accident Studies Christopher S. Bajwa Division of Spent Fuel Storage and Transportation Office of Nuclear Material Safety and Safeguards USNRC 2012 U.S. DOE National Transportation Stakeholders Forum (NTSF) May 15 - 17, 2012 Knoxville, TN * Going The Distance? - The Safe Transport of Spent Nuclear Fuel and High-Level Radioactive Waste in the United States * Released February 9, 2006 * Conclusions: * NRC safety regulations are adequate to ensure package containment effectiveness over a wide range of transport conditions, including most credible accident conditions. * The radiological risks are well understood and are generally low, with the possible exception of risks from releases in extreme accidents involving long duration, fully engulfing fires.

97

Accident resistant transport container  

DOE Patents (OSTI)

The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

Andersen, John A. (Albuquerque, NM); Cole, James K. (Albuquerque, NM)

1980-01-01T23:59:59.000Z

98

Analysis of a Nuclear Accident: Fission and Activation Product Releases from the Fukushima Daiichi Nuclear Facility as Remote Indicators of Source Identification, Extent of Release, and State of Damaged Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

Measurements of several radionuclides within environmental samples taken from the Fukushima Daiichi nuclear facility and reported on the Tokyo Electric Power Company website following the recent tsunami-initiated catastrophe were evaluated for the purpose of identifying the source term, reconstructing the release mechanisms, and estimating the extent of the release. 136Cs/137Cs and 134Cs/137Cs ratios identified Units 1-3 as the major source of radioactive contamination to the surface soil close to the facility. A trend was observed between the fraction of the total core inventory released for a number of fission product isotopes and their corresponding Gibbs Free Energy of formation for the primary oxide form of the isotope, suggesting that release was dictated primarily by chemical volatility driven by temperature and reduction potential within the primary containment vessels of the vented reactors. The absence of any major fractionation beyond volatilization suggested all coolant had evaporated by the time of venting. High estimates for the fraction of the total inventory released of more volatile species (Te, Cs, I) indicated the damage to fuel bundles was likely extensive, minimizing any potential containment due to physical migration of these species through the fuel matrix and across the cladding wall. 238Pu/239,240Pu ratios close-in and at 30 km from the facility indicated that the damaged reactors were the major contributor of Pu to surface soil at the source but that this contribution likely decreased rapidly with distance from the facility. The fraction of the total Pu inventory released to the environment from venting units 1 and 3 was estimated to be ~0.003% based upon Pu/Cs isotope ratios relative to the within-reactor modeled inventory prior to venting and was consistent with an independent model evaluation that considered chemical volatility based upon measured fission product release trends. Significant volatile radionuclides within the spent fuel at the time of venting but not as yet observed and reported within environmental samples are suggested as potential analytes of concern for future environmental surveys around the site.

Schwantes, Jon M.; Orton, Christopher R.; Clark, Richard A.

2012-09-10T23:59:59.000Z

99

MOSES Leak Tool 1.0 - Mineral Oil Spill Evaluation System Leak Tool, Version 1.0  

Science Conference Proceedings (OSTI)

The purpose of the Mineral Oil Spill Evaluation System (MOSES) Leak Tool Version 1.0 is to provide a Monte-Carlo estimate of the initial horizontal spill radius from leaks in either at-grade or pole-mounted transformers. The internal transformer pressure is specified as either being atmospheric or at pressurized conditions. This tool is intended to supplement the MOSES-MP code (EPRI, 2002). The MOSES-MP code estimates the extent of oil migration from leaks and spills from electrical oil-filled equipment ...

2007-08-20T23:59:59.000Z

100

MAAP5 Simulation of Accidents at Fukushima Dai-ichi Units 1, 2, and 3  

Science Conference Proceedings (OSTI)

The original MAAP4 code functional design specification (circa 1989) was defined to address the full extent of degraded core accidents with the potential for reflooding of a badly damaged core. It was intended to support probabilistic risk assessment (PRA) and severe accident management guideline (SAMG) applications that previously were limited by the relatively rudimentary design for MAAP3.0B, the predecessor code.The accidents at Fukushima Dai-ichi Units 1, 2, and 3 prompted a ...

2013-02-23T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Statistical approaches to leak detection for geological sequestration  

E-Print Network (OSTI)

Geological sequestration has been proposed as a way to remove CO? from the atmosphere by injecting it into deep saline aquifers. Detecting leaks to the atmosphere will be important for ensuring safety and effectiveness of ...

Haidari, Arman S

2011-01-01T23:59:59.000Z

102

Using Cyclic Memory Allocation to Eliminate Memory Leaks  

E-Print Network (OSTI)

We present and evaluate a new memory management technique for eliminating memory leaks in programs with dynamic memory allocation. This technique observes the execution of the program on a sequence of training inputs to ...

Nguyen, Huu Hai

103

Using Cyclic Memory Allocation to Eliminate Memory Leaks  

E-Print Network (OSTI)

We present and evaluate a new memory management technique foreliminating memory leaks in programs with dynamic memoryallocation. This technique observes the execution of the program on asequence of training inputsto find ...

Nguyen, Huu Hai

2005-10-26T23:59:59.000Z

104

Predicting Worker Exposure from a Glovebox Leak  

Science Conference Proceedings (OSTI)

It is difficult to predict immediate worker radiological consequences from a hypothetical accident. This is recognized in DOE safety analysis guidance and the reason such guidance does not call for quantitative determinations of such consequences. However, it would be useful to at least have a means of systematically and formally quantifying worker dose to be able to identify the relative risks of various processes and to provide an order-of-magnitude impression of absolute consequences. In this report, we present such a means in the form of a simple calculation model that is easily applied and generates reasonable, qualitative dose predictions. The model contains a scaling parameter whose value was deduced from extensive laboratory ventilation flow rate measurements performed at Los Alamos National Laboratory (LANL) over the last several years and from recent indoor radioactive contamination dispersion measurements, also at LANL. Application of the model is illustrated with the aid of two example calculations.

H. Jordan; D. J. Gordon; J. J. Whicker; D. L. Wannigman

2001-05-01T23:59:59.000Z

105

Leak test fixture and method for using same  

DOE Patents (OSTI)

A method and apparatus are provided which are especially useful for leak testing seams such as an end closure or joint in an article. The test does not require an enclosed pressurized volume within the article or joint section to be leak checked. A flexible impervious membrane is disposed over an area of the seamed surfaces to be leak checked and sealed around the outer edges. A preselected vacuum is applied through an opening in the membrane to evacuate the area between the membrane and the surface being leak checked to essentially collapse the membrane to conform to the article surface or joined adjacent surfaces. A pressure differential is concentrated at the seam bounded by the membrane and only the seam experiences a pressure differential as air or helium molecules are drawn into the vacuum system through a leak in the seam. A helium detector may be placed in a vacuum exhaust line from the membrane to detect the helium. Alternatively, the vacuum system may be isolated at a preselected pressure and leaks may be detected by a subsequent pressure increase in the vacuum system.

Hawk, Lawrence S. (Knoxville, TN)

1976-01-01T23:59:59.000Z

106

Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.  

Science Conference Proceedings (OSTI)

An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

2010-03-01T23:59:59.000Z

107

Accident Investigation Handbook  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SENSI NOT MEAS UREMENT TIVE D DOE-HDBK-1 1208-2012 July 2012 DOE E HA ANDBOOK K Ac ccide ent and d Op pera ational Sa afety y An naly ysis Volume e I: Ac ccide ent A Analy ysis Tec chniq ques U.S. Depar rtmen nt of En nergy Was shingto on, D.C C. 205 85 DOE-HDBK-1208-2012 INTRODUCTION - HANDBOOK APPLICATION AND SCOPE Accident Investigations (AI) and Operational Safety Reviews (OSR) are valuable for evaluating technical issues, safety management systems and human performance and environmental conditions to prevent accidents, through a process of continuous organizational learning. This Handbook brings together the strengths of the experiences gained in conducting Department of Energy (DOE) accident investigations over the past many years. That experience encourages us

108

Microsoft Word - Unrelated Accident  

NLE Websites -- All DOE Office Websites (Extended Search)

For Immediate Release For Immediate Release Truck Accident Did Not Involve WIPP Shipment CARLSBAD, N.M., October 1, 2009 - A Wednesday night truck accident north of Albuquerque on Highway 165 that involved an 18-wheeler is not related to Waste Isolation Pilot Plant (WIPP) transuranic waste shipments. Involved in the accident was a load of new, unused 55-gallon drums manufactured in Carlsbad that was en route to Richland, Washington. The Waste Isolation Pilot Plant is a U.S. Department of Energy facility designed to safely isolate defense-related transuranic waste from people and the environment. Waste temporarily stored at sites around the country is shipped to WIPP and permanently disposed in rooms mined out of an ancient salt formation 2,150 feet below the surface. WIPP, which began waste

109

Commercial SNF Accident Release Fractions  

Science Conference Proceedings (OSTI)

The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the container that confines the fuel assemblies could provide an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. This analysis, however, does not take credit for the additional barrier and establishes only the total release fractions for bare unconfined intact commercial SNF assemblies, which may be conservatively applied to confined intact commercial I SNF assemblies.

J. Schulz

2004-11-05T23:59:59.000Z

110

Computerized Accident Incident Reporting System  

Energy.gov (U.S. Department of Energy (DOE))

The Computerized Accident/Incident Reporting System is a database used to collect and analyze DOE and DOE contractor reports of injuries, illnesses, and other accidents that occur during DOE...

111

Prediction of Gas Leak Tightness of Superplastically Formed Products  

Science Conference Proceedings (OSTI)

In some applications, in this case an aluminium box in a subatomic particle detector containing highly sensitive detecting devices, it is important that a formed sheet should show no gas leak from one side to the other. In order to prevent a trial-and-error procedure to make this leak tight box, a method is set up to predict if a formed sheet conforms to the maximum leak constraint. The technique of superplastic forming (SPF) is used in order to attain very high plastic strains before failure. Since only a few of these boxes are needed, this makes, this generally slow, process an attractive production method. To predict the gas leak of a superplastically formed aluminium sheet in an accurate way, finite element simulations are used in combination with a user-defined material model. This constitutive model couples the leak rate with the void volume fraction. This void volume fraction is then dependent on both the equivalent plastic strain and the applied hydrostatic pressure during the bulge process (backpressure).

Snippe, Corijn H. C. [National Institute for Subatomic Physics (Nikhef) PO Box 41882, 1009 DB Amsterdam (Netherlands); Meinders, T. [University of Twente, Faculty of Engineering Technology PO Box 217, 7500 AE Enschede (Netherlands)

2010-06-15T23:59:59.000Z

112

WHEN MODEL MEETS REALITY A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT  

SciTech Connect

The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significant damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the real accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.

Zhegang Ma

2013-09-01T23:59:59.000Z

113

Oil/gas collector/separator for underwater oil leaks  

DOE Patents (OSTI)

This invention is comprised of an oil/gas collector/separator for recovery of oil leaking, for example, from an offshore or underwater oil well. The separator is floated over the point of the leak and tethered in place so as to receive oil/gas floating, or forced under pressure, toward the water surface from either a broken or leaking oil well casing, line, or sunken ship. The separator is provided with a downwardly extending skirt to contain the oil/gas which floats or is forced upward into a dome wherein the gas is separated from the oil/water, with the gas being flared (burned) at the top of the dome, and the oil is separated from water and pumped to a point of use. Since the density of oil is less than that of water it can be easily separated from any water entering the dome.

Henning, C.D.

1992-12-31T23:59:59.000Z

114

Oil/gas collector/separator for underwater oil leaks  

DOE Patents (OSTI)

An oil/gas collector/separator for recovery of oil leaking, for example, from an offshore or underwater oil well. The separator is floated over the point of the leak and tethered in place so as to receive oil/gas floating, or forced under pressure, toward the water surface from either a broken or leaking oil well casing, line, or sunken ship. The separator is provided with a downwardly extending skirt to contain the oil/gas which floats or is forced upward into a dome wherein the gas is separated from the oil/water, with the gas being flared (burned) at the top of the dome, and the oil is separated from water and pumped to a point of use. Since the density of oil is less than that of water it can be easily separated from any water entering the dome.

Henning, Carl D. (Livermore, CA)

1993-01-01T23:59:59.000Z

115

Apparatus and method for detecting leaks in piping  

DOE Patents (OSTI)

A method and device are disclosed for detecting the location of leaks along a wall or piping system, preferably in double-walled piping. The apparatus comprises a sniffer probe, a rigid cord such as a length of tube attached to the probe on one end and extending out of the piping with the other end, a source of pressurized air and a source of helium. The method comprises guiding the sniffer probe into the inner pipe to its distal end, purging the inner pipe with pressurized air, filling the annulus defined between the inner and outer pipe with helium, and then detecting the presence of helium within the inner pipe with the probe as is pulled back through the inner pipe. The length of the tube at the point where a leak is detected determines the location of the leak in the pipe. 2 figures.

Trapp, D.J.

1994-12-27T23:59:59.000Z

116

A comparative analysis of accident risks in fossil, hydro, and nuclear energy chains  

Science Conference Proceedings (OSTI)

This study presents a comparative assessment of severe accident risks in the energy sector, based on the historical experience of fossil (coal, oil, natural gas, and LPG (Liquefied Petroleum Gas)) and hydro chains contained in the comprehensive Energy-related Severe Accident Database (ENSAD), as well as Probabilistic Safety Assessment (PSA) for the nuclear chain. Full energy chains were considered because accidents can take place at every stage of the chain. Comparative analyses for the years 1969-2000 included a total of 1870 severe ({>=} 5 fatalities) accidents, amounting to 81,258 fatalities. Although 79.1% of all accidents and 88.9% of associated fatalities occurred in less developed, non-OECD countries, industrialized OECD countries dominated insured losses (78.0%), reflecting their substantially higher insurance density and stricter safety regulations. Aggregated indicators and frequency-consequence (F-N) curves showed that energy-related accident risks in non-OECD countries are distinctly higher than in OECD countries. Hydropower in non-OECD countries and upstream stages within fossil energy chains are most accident-prone. Expected fatality rates are lowest for Western hydropower and nuclear power plants; however, the maximum credible consequences can be very large. Total economic damages due to severe accidents are substantial, but small when compared with natural disasters. Similarly, external costs associated with severe accidents are generally much smaller than monetized damages caused by air pollution.

Burgherr, P.; Hirschberg, S. [Paul Scherrer Institute, Villigen (Switzerland)

2008-07-01T23:59:59.000Z

117

High Altitude Aerial Natural Gas Leak Detection System  

SciTech Connect

The objective of this program was to develop and demonstrate a cost-effective and power-efficient advanced standoff sensing technology able to detect and quantify, from a high-altitude (> 10,000 ft) aircraft, natural gas leaking from a high-pressure pipeline. The advanced technology is based on an enhanced version of the Remote Methane Leak Detector (RMLD) platform developed previously by Physical Sciences Inc. (PSI). The RMLD combines a telecommunications-style diode laser, fiber-optic components, and low-cost DSP electronics with the well-understood principles of Wavelength Modulation Spectroscopy (WMS), to indicate the presence of natural gas located between the operator and a topographic target. The transceiver transmits a laser beam onto a topographic target and receives some of the laser light reflected by the target. The controller processes the received light signal to deduce the amount of methane in the laser's path. For use in the airborne platform, we modified three aspects of the RMLD, by: (1) inserting an Erbium-doped optical fiber laser amplifier to increase the transmitted laser power from 10 mW to 5W; (2) increasing the optical receiver diameter from 10 cm to 25 cm; and (3) altering the laser wavelength from 1653 nm to 1618 nm. The modified RMLD system provides a path-integrated methane concentration sensitivity {approx}5000 ppm-m, sufficient to detect the presence of a leak from a high capacity transmission line while discriminating against attenuation by ambient methane. In ground-based simulations of the aerial leak detection scenario, we demonstrated the ability to measure methane leaks within the laser beam path when it illuminates a topographic target 2000 m away. We also demonstrated simulated leak detection from ranges of 200 m using the 25 cm optical receiver without the fiber amplifier.

Richard T. Wainner; Mickey B. Frish; B. David Green; Matthew C. Laderer; Mark G. Allen; Joseph R. Morency

2006-12-31T23:59:59.000Z

118

Barriers to Switching Accidents  

Science Conference Proceedings (OSTI)

The EPRI Switching Safety & Reliability Project Steering Committee sponsored development of a self-study based training program for personnel who perform switching. Some of the earlier EPRI Switching Safety & Reliability research projects that focused on the causes of switching errors, highlighted a need to reduce the 'complacency' that tends to develop as switching activities are performed over and over again and become 'routine.' Most switching accidents or incidents involve personnel who were trained ...

2005-12-22T23:59:59.000Z

119

New concepts for refrigerant leak detection and mixture measurement  

Science Conference Proceedings (OSTI)

Since the discovery that chlorofluorocarbons (CFCs) destroy the ozone layer, the need to reduce the release of these refrigerants into the environment has become critical. A total ban of ozone-depleting CFCs is expected within a few years, and hydrofluorocarbons (HFCs) and fluorocarbons (FCs) and their mixtures are expected to be used during a transition period. Several HFC and FC refrigerants are currently being considered as CFC substitutes. The electronic refrigerant leak detectors currently being considered as CFC substitutes. The electronic refrigerant leak detectors currently on the market were developed to detect CFCs and are not as sensitive to HFCs. Although incremental improvement can be made to these devices to detect HFCs, they often lead to increased false signals. Also, there is no simple device available to measure the composition of a refrigerant mixture. The authors present two new concepts to aid in the development of two portable instruments that can be used for HFC leak detection and for quantitative measurement of refrigerant mixture compositions. The development of simple, easy-to-use portable leak detectors and refrigerant mixture meters is essential to the wide use of alternative refrigerants in industry.

Chen, F.C.; Allman, S.L.; Chen, C.H.

1993-12-31T23:59:59.000Z

120

How Do You Find Thermal Leaks in Your Home? | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

How Do You Find Thermal Leaks in Your Home? How Do You Find Thermal Leaks in Your Home? How Do You Find Thermal Leaks in Your Home? March 31, 2011 - 7:30am Addthis On Monday, John told you about the thermal leak detector he purchased to help him find and seal leaks in his home. A thermal leak detector can be a great tool to help you find leaks in your own home, but it's not your only option. In addition to tools like this, you can also use some of our tips on do-it-yourself energy assessments, or you could get a professional energy assessment. How do you find thermal leaks in your home? Each Thursday, you have the chance to share your thoughts on a question about energy efficiency or renewable energy for consumers. Please e-mail your responses to the Energy Saver team at consumer.webmaster@nrel.gov.

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010 (2 of 4) BP Oil Spill Footage (High Def) - Leak at 4840' - June 3 2010 (1 of 4) Re-Building Greensburg The...

122

BP Oil Spill Footage (High Def) - Leak at 4840' - June 3 2010...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010 (3 of 4) BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010 (2 of 4) Re-Building Greensburg The...

123

BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010 (3 of 4) BP Oil Spill Footage (High Def) - Leak at 4840' - June 3 2010 (1 of 4) Re-Building Greensburg The...

124

Analysis and design of an in-pipe system for water leak detection  

E-Print Network (OSTI)

Leaks are a major factor for unaccounted water losses in almost every water distribution network. Pipeline leak may result, for example, from bad workmanship or from any destructive cause, due to sudden changes of pressure, ...

Chatzigeorgiou, Dimitris M

2010-01-01T23:59:59.000Z

125

CSNI specialist meeting on leak-before-break in nuclear reactor piping: proceedings  

SciTech Connect

On September 1 and 2, 1983, the CSNI subcommittee on primary system integrity held a special meeting in Monterey, California, on the subject of leak-before-break in nuclear reactor piping systems. The purpose of the meeting was to provide an international forum for the exchange of ideas, positions, and research results; to identify areas requiring additional research and development; and to determine the general attitude toward acceptance of the leak-before-break concept. The importance of the leak-before-break issue was evidenced by excellent attendance at the meeting and through active participation by the meeting attendees. Approximately 125 people representing fifteen different nations attended the meeting. The meeting was divided into four technical sessions addressing the following areas: Application of Piping Fracture Mechanics to Leak-Before Break, Leak Rate and Leak Detection, Leak-Before-Break Studies, Methods and Results, Current and Proposed Positions on Leak-Before-Break.

Not Available

1984-08-01T23:59:59.000Z

126

Methodology to quantify leaks in aerosol sampling system components  

E-Print Network (OSTI)

Filter holders and continuous air monitors (CAMs) are used extensively in the nuclear industry. It is important to minimize leakage in these devices and in recognition of this consideration, a limit on leakage for sampling systems is specified in ANSI/HPS N13.1-1999; however the protocol given in the standard is really germane to measurement of significant leakage, e.g., several percent of the sampling flow rate. In the present study, a technique for quantifying leakage was developed and that approach was used to measure the sealing integrity of a CAM and two kinds of filter holders. The methodology involves use of sulfur hexafluoride as a tracer gas with the device being tested operated under dynamic flow conditions. The leak rates in these devices were determined in the pressure range from 2.49 kPa (10 In. H2O) vacuum to 2.49 kPa (10 In. H2O) pressure at a typical flow rate of 56.6 L/min (2 cfm). For the two filter holders, the leak rates were less than 0.007% of the nominal flow rate. The leak rate in the CAM was less than 0.2% of the nominal flow rate. These values are well within the limit prescribed in the ANSI standard, which is 5% of the nominal flow rate. Therefore the limit listed in the ANSI standard should be reconsidered as lower values can be achieved, and the methodology presented herein can be used to quantify lower leakage values in sample collectors and analyzers. A theoretical analysis was also done to determine the nature of flow through the leaks and the amount of flow contribution by the different possible mechanisms of flow through leaks.

Vijayaraghavan, Vishnu Karthik

2003-08-01T23:59:59.000Z

127

Severe accident analysis using dynamic accident progression event trees.  

E-Print Network (OSTI)

??In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce (more)

Hakobyan, Aram P

2006-01-01T23:59:59.000Z

128

Hanford Single-Shell Tank Leak Causes and Locations - 241-B Farm  

SciTech Connect

This document identifies 241-B Tank Farm (B Farm) leak cause and locations for the 100 series leaking tank (241-B-107) identified in RPP-RPT-49089, Hanford B-Farm Leak Inventory Assessments Report. This document satisfies the B Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

Girardot, Crystal L. [Washington River Protection Systems, Richland, WA (United States); Harlow, Donald G. [Washington River Protection Systems, Richland, WA (United States)

2013-07-11T23:59:59.000Z

129

Stress in accident and post-accident management at Chernobyl ?  

E-Print Network (OSTI)

Abstract. The effects of the Chernobyl nuclear accident on the psychology of the affected population have been much discussed. The psychological dimension has been advanced as a factor explaining the emergence, from 1990 onwards, of a post-accident crisis in the main CIS countries affected. This article presents the conclusions of a series of European studies, which focused on the consequences of the Chernobyl accident. These studies show that the psychological and social effects associated with the post-accident situation arise from the interdependency of a number of complex factors exerting a deleterious effect on the population. We shall first attempt to characterise the stress phenomena observed among the population affected by the accident. Secondly, we will be presenting an analysis of the various factors that have contributed to the emerging psychological and social features of population reaction to the accident and in post-accident phases, while not neglecting the effects of the pre-accident situation on the target population. Thirdly, we shall devote some initial consideration to the conditions that might be conducive to better management of postaccident stress. In conclusion, we shall emphasise the need to restore confidence among the population generally. 1.

Gilles Heriard Dubreuil

1996-01-01T23:59:59.000Z

130

Method and means of passive detection of leaks in buried pipes  

DOE Patents (OSTI)

A method and means for passive detection of a leak in a buried pipe containing fluid under pressure includes a plurality of acoustic detectors that are placed in contact with the pipe. Noise produced by the leak is detected by the detectors, and the detected signals are correlated to locate the leak. In one embodiment of the invention two detectors are placed at different locations to locate a leak between them. In an alternate embodiment two detectors of different waves are placed at substantially the same location to determine the distance of the leak from the location.

Claytor, T.

1979-10-30T23:59:59.000Z

131

Comments on the leak-before-break concept for nuclear power plant piping systems  

SciTech Connect

The leak-before-break concept is based on the idea that, with a high degree of probability, failure of the pressure boundary of piping systems will be signaled by a detectable leak that will provide ample time to shutdown and repair that leak. The status of the leak-before-break concept is discussed in this report, including a review of industrial and nuclear power plant experience with respect to leak-before-break, fracture mechanics, and potential elimination of postulated pipe breaks in nuclear power plant piping design. 36 refs., 12 figs., 3 tabs.

Rodabaugh, E.C.

1985-08-01T23:59:59.000Z

132

Method and means of passive detection of leaks in buried pipes  

DOE Patents (OSTI)

A method and means for passive detection of a leak in a buried pipe containing fluid under pressure includes a plurality of acoustic detectors that are placed in contact with the pipe. Noise produced by the leak is detected by the detectors, and the detected signals are correlated to locate the leak. In one embodiment of the invention two detectors are placed at different locations to locate a leak between them. In an alternate embodiment two detectors of different waves are placed at substantially the same location to determine the distance of the leak from the location.

Claytor, Thomas N. (Woodridge, IL)

1981-01-01T23:59:59.000Z

133

Leake County, Mississippi: Energy Resources | Open Energy Information  

Open Energy Info (EERE)

Leake County, Mississippi: Energy Resources Leake County, Mississippi: Energy Resources Jump to: navigation, search Equivalent URI DBpedia Coordinates 32.8073509°, -89.4742177° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":32.8073509,"lon":-89.4742177,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

134

Distributed Optical Sensor for CO2 Leak Detection  

NLE Websites -- All DOE Office Websites (Extended Search)

Optical Sensor for CO Optical Sensor for CO 2 Leak Detection Opportunity Research is active on the technology "Distributed Optical Sensor for CO 2 Leak Detection," for which a Patent Application has been filed. This technology is available for licensing and/or further collaborative research from the U.S. Department of Energy's National Energy Technology Laboratory (NETL). Overview The availability of fossil fuels to provide clean, affordable energy is essential for domestic and global prosperity and security well into the 21st century. However, there are concerns over the impacts of greenhouse gases (GHGs) in the atmosphere-particularly carbon dioxide (CO 2 ). Carbon capture and storage in geologic formations is a promising technology to reduce the impact of CO

135

Management of Leaks in Hydrogen Production, Delivery, and Storage Systems  

DOE Green Energy (OSTI)

A systematic approach to manage hydrogen leakage from components is presented. Methods to evaluate the quantity of hydrogen leakage and permeation from a system are provided by calculation and testing sensitivities. The following technology components of a leak management program are described: (1) Methods to evaluate hydrogen gas loss through leaks; (2) Methods to calculate opening areas of crack like defects; (3) Permeation of hydrogen through metallic piping; (4) Code requirements for acceptable flammability limits; (5) Methods to detect flammable gas; (6) Requirements for adequate ventilation in the vicinity of the hydrogen system; (7) Methods to calculate dilution air requirements for flammable gas mixtures; and (8) Concepts for reduced leakage component selection and permeation barriers.

Rawls, G

2006-04-27T23:59:59.000Z

136

MCO combustible gas management leak test acceptance criteria  

DOE Green Energy (OSTI)

Existing leak test acceptance criteria for mechanically sealed and weld sealed multi-canister overpacks (MCO) were evaluated to ensure that MCOs can be handled and stored in stagnant air without compromising the Spent Nuclear Fuel Project's overall strategy to prevent accumulation of combustible gas mixtures within MCO's or within their surroundings. The document concludes that the integrated leak test acceptance criteria for mechanically sealed and weld sealed MCOs (1 x 10{sup -5} std cc/sec and 1 x 10{sup -7} std cc/sec, respectively) are adequate to meet all current and foreseeable needs of the project, including capability to demonstrate compliance with the NFPA 60 Paragraph 3-3 requirement to maintain hydrogen concentrations [within the air atmosphere CSB tubes] t or below 1 vol% (i.e., at or below 25% of the LFL).

SHERRELL, D.L.

1999-05-11T23:59:59.000Z

137

Cambium Damage  

NLE Websites -- All DOE Office Websites (Extended Search)

Cambium Damage Cambium Damage Name: Jamie Location: N/A Country: N/A Date: N/A Question: If the bark from the lower part of trees (elm trees) is almost completly removed (in this case by animals)to a height of about 8ft, is it possible that the trees will still live? What can be done to help the trees? Replies: If the tree has been girdled, that is, the bark and cambium layer beneath it, has been removed completely around the tree, then it will die. If there is any portion of the bark remaining it may live, but if that remaining is small it probably will die fairly soon due to general decline. If the cambium layer has not been destroyed it may recover, but once the bark is stripped away it is most likely doomed because of the likelihood of invasion by fungi, insects, etc. A local forester or landscaper might be able to offer more help if they see it.

138

Accurate accident reconstruction in VANET  

Science Conference Proceedings (OSTI)

We propose a forensic VANET application to aid an accurate accident reconstruction. Our application provides a new source of objective real-time data impossible to collect using existing methods. By leveraging inter-vehicle communications, we compile ... Keywords: EDR, VANET, accident reconstruction, in-vehicle applications

Yuliya Kopylova; Csilla Farkas; Wenyuan Xu

2011-07-01T23:59:59.000Z

139

Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals  

Science Conference Proceedings (OSTI)

This report presents a risk impact assessment for extending integrated leak rate test (ILRT) surveillance intervals to 15 years. The assessment demonstrates that on an industry-wide basis there is small risk associated with the extension, provided that the performance bases and defense-in-depth are maintained. There is an obvious benefit in not performing costly, critical-path, time-consuming tests that provide a limited benefit from a risk perspective.

2008-10-31T23:59:59.000Z

140

Demonstration of KEMA SF6 Leak Detector at Consolidated Edison  

Science Conference Proceedings (OSTI)

Detecting leaks of sulfur hexafluoride (SF6) gas has become more important as environmental regulators forge ahead with programs aimed at curbing SF6 emissions and energy companies seek to cut costs. SF6 is widely used in the electric power industry as an insulator for high-voltage circuit breakers, switchgear, and other substation equipment. A new on-line applicable SF6 leakage detection technique (KEMA patented) using photo-acoustic detection of SF6 was researched during demonstrations at Consolidated ...

2006-09-18T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Analysis of the leak-detection system for top welds of EBR-II fuel elements  

SciTech Connect

An analysis of the leak detector used to check the top welds on EBR-II fuel elements was performed. Data were obtained to allow calculation of volumes of the metering chamber and test chamber at each station of the leak detector. These volumes and a mathematical model were used to calculate decrease in pressure with time for each station. Values for calibrated leaks and unknown leak rates were compared with calculated ones. The calculated results for the two calibrated leaks agreed with the observed pressure-time results for the two leaks. Results show that determining the volumes of each leak-detector station allows the leakrate sensitivity to be readily calculated for each station. One leak-detector station could not detect a minimum leak rate of 2 x 10/sup -4/ std cm/sup 3/ sec at 40 atm, which is the current specification. The other four stations could meet the specification. Suggestions are given for periodic calibration of the leak detectors as well as precautions that must be observed to achieve optimum sensitivity when operating the leak detector. (auth)

Hudman, G.D.; Walters, L.C.

1973-05-01T23:59:59.000Z

142

MWTF jumper connector integral seal block development and leak testing  

SciTech Connect

In fiscal year 1993, tests of an o-ring/tetraseal retainer designed to replace a gasket-type seal used in PUREX-type process jumper connectors encouraged the design of an improved seal block. This new seal block combines several parts into one unitized component called an integral seal block. This report summarizes development and leak testing of the new integral seal block. The integral seal block uses a standard o-ring nested in a groove to accomplish leak tightness. This seal block eliminates the need to machine acme threads into the lower skirt casting and seal retainers, eliminates tolerance stack-up, reduces parts inventory, and eliminates an unnecessary leak path in the jumper connector assembly. This report also includes test data on various types of o-ring materials subjected to heat and pressure. Materials tested included Viton, Kalrez, and fluorosilicone, with some incidental data on teflon coated silicone o-rings. Test experience clearly demonstrates the need to test each seal material for temperature and pressure in its intended application. Some materials advertised as being {open_quotes}better{close_quotes} at higher temperatures did not perform up to expectations. Inspection of the fluorosilicone and Kalrez seals after thermal testing indicates that they are much more susceptible to heat softening than Viton.

Ruff, E.S.; Jordan, S.R.

1995-01-01T23:59:59.000Z

143

Calculation of SY tank annulus continuous air monitor readings after postulated leak scenarios  

Science Conference Proceedings (OSTI)

The objective of this work was to determine whether or not a continuous air monitor (CAM) monitoring the annulus of one of the SY Tanks would be expected to alarm after three postulated leak scenarios. Using data and references provided by Lockheed Martin`s Tank Farm personnel, estimated CAM readings were calculated at specific times after the postulated scenarios might have occurred. Potential CAM readings above background at different times were calculated for the following leak scenarios: Leak rate of 0.01 gal/min; Leak rate of 0.03 gal/min (best estimate of the maximum probable leak rate from a single-shell tank); and Leak of 73 gal (equivalent to a {1/4}-in. leak on the floor of the annulus). The equation used to make the calculations along with descriptions and/or explanations of the terms are included, as is a list of the assumptions and/or values used for the calculations.

Kenoyer, J.L.

1998-08-01T23:59:59.000Z

144

Rail transportation risk and accident severity: A statistical analysis of variables in FRA's accident/incident data base  

Science Conference Proceedings (OSTI)

The Federal Railroad Administration (US DOT) maintains a file of carrier-reported railroad accidents and incidents that meet stipulated threshold criteria for damage cost and/or casualties. A thoroughly-cleaned five-year time series of this data base was subjected to unbiased statistical procedures to discover (a) important causative variables in severe (high damage cost) accidents and (b) other key relationships between objective accident conditions and frequencies. Just under 6000 records, each representing a single event involving rail freight shipments moving on mainline track, were subjected to statistical frequency analysis, then included in the construction of classification and regression trees as described by Breimann et al. (1984). Variables related to damage cost defined the initial splits,'' or branchings of the tree. An interesting implication of the results of this analysis with respect to transportation of hazardous wastes by rail is that movements should be avoided when ambient temperatures are extreme (significantly 80{degrees}F), but that there should be no a priori bias against shipping wastes in longer train consists. 2 refs., 2 figs., 12 tabs.

Saricks, C.L. (Argonne National Lab., IL (USA). Energy Systems Div.); Janssen, I. (Argonne National Lab., IL (USA). Biological and Medical Research Div.)

1991-01-01T23:59:59.000Z

145

Property Damage Risk Assessment Scoping Study: for South Texas Project Electric Generating Station  

Science Conference Proceedings (OSTI)

At the request of the South Texas Project Electric Generating Station (STPEGS), EPRI assessed the financial risks of on-site property damage from component failures and accidents and the effectiveness of available insurance in mitigating such risks. This report quantifies the risks of nuclear and nonnuclear accidents and the resulting property damage incurred. The report is a companion document to EPRI's Nuclear Property Insurance Study (TR-108061), which discusses five options for alternate insurance co...

1997-08-12T23:59:59.000Z

146

Purify: Fast Detection of Memory Leaks and Access Errors This paper describes Purify, a software testing and quality assurance tool that detects memory leaks and  

E-Print Network (OSTI)

], Catalytix [Feuer85] and various similar malloc_debug packages use. Byte and two-byte checking cannot spent eliminating leaks in the X11R4 server for Sun workstations. All that effort, yet dozens of leaks as gcc. The data was collected on a Sun SPARCstation SLC running SUNOS 4.1.1, and all times are real

Qin, Feng

147

Properly designed underbalanced drilling fluids can limit formation damage  

Science Conference Proceedings (OSTI)

Drilling fluids for underbalanced operations require careful design and testing to ensure they do not damage sensitive formations. In addition to hole cleaning and lubrication functions, these fluids may be needed as kill fluids during emergencies. PanCanadian Petroleum Ltd. used a systematic approach in developing and field testing a nondamaging drilling fluid. It was for use in underbalanced operations in the Glauconitic sandstone in the Westerose gas field in Alberta. A lab study was initiated to develop and test a non-damaging water-based drilling fluid for the horizontal well pilot project. The need to develop an inexpensive, nondamaging drilling fluid was previously identified during underbalanced drilling operations in the Weyburn field in southeastern Saskatchewan. A non-damaging fluid is required for hole cleaning, for lubrication of the mud motor, and for use as a kill fluid during emergencies. In addition, a nondamaging fluid is required when drilling with a conventional rig because pressure surges during connections and trips may result in the well being exposed to short periods of near balanced or overbalanced conditions. Without the protection of a filter cake, the drilling fluid will leak off into the formation, causing damage. The amount of damage is related to the rate of leak off and depth of invasion, which are directly proportional to the permeability to the fluid.

Churcher, P.L.; Yurkiw, F.J. [PanCanadian Petroleum Ltd., Calgary, Alberta (Canada); Bietz, R.F.; Bennion, D.B. [Hycal Energy Research Ltd., Calgary, Alberta (Canada)

1996-04-29T23:59:59.000Z

148

OSSA - An optimized approach to severe accident management: EPR application  

SciTech Connect

There is a recognized need to provide nuclear power plant technical staff with structured guidance for response to a potential severe accident condition involving core damage and potential release of fission products to the environment. Over the past ten years, many plants worldwide have implemented such guidance for their emergency technical support center teams either by following one of the generic approaches, or by developing fully independent approaches. There are many lessons to be learned from the experience of the past decade, in developing, implementing, and validating severe accident management guidance. Also, though numerous basic approaches exist which share common principles, there are differences in the methodology and application of the guidelines. AREVA/Framatome-ANP is developing an optimized approach to severe accident management guidance in a project called OSSA ('Operating Strategies for Severe Accidents'). There are still numerous operating power plants which have yet to implement severe accident management programs. For these, the option to use an updated approach which makes full use of lessons learned and experience, is seen as a major advantage. Very few of the current approaches covers all operating plant states, including shutdown states with the primary system closed and open. Although it is not necessary to develop an entirely new approach in order to add this capability, the opportunity has been taken to develop revised full scope guidance covering all plant states in addition to the fuel in the fuel building. The EPR includes at the design phase systems and measures to minimize the risk of severe accident and to mitigate such potential scenarios. This presents a difference in comparison with existing plant, for which severe accidents where not considered in the design. Thought developed for all type of plants, OSSA will also be applied on the EPR, with adaptations designed to take into account its favourable situation in that field. This revised approach will incorporate a number of new features which will simplify and streamline the guidance material while ensuring comprehensive guidance for response to any severe accident. Examples of such features include : - Identification of severe accident challenges based on plant specific studies. - Revision of the split of responsibilities between operations and technical support center staff. - Fixed setpoint entry conditions, ensuring that the transition from emergency procedures takes place at a consistent core/fuel condition (regardless of scenario), and which fixes the time window available to attempt ultimate preventive measures. - A safety function concept for monitoring plant conditions (in the control room). - An integrated graphic-based diagnostic tool including entry condition, challenge prioritization, and exit condition monitoring to be used by the technical support team. This paper describes the basic features of OSSA, and project status. (authors)

Sauvage, E. C.; Prior, R.; Coffey, K. [AREVA, FRAMATOME-ANP SAS, Paris, 92084 La Defense (France); Mazurkiewicz, S. M. [AREVA, FRAMATOME-ANP Inc, Lynchburg, VA 24506-0935 (United States)

2006-07-01T23:59:59.000Z

149

Bottom-Fill Method for Stopping Leaking Oil Wells  

E-Print Network (OSTI)

Hardware failure at the top of a deep underwater oil well can result in a catastrophic oil leak. The enormous pressure lifting the column of oil in that well makes it nearly impossible to stop from the top with seals or pressurization. We propose to fill the bottom of the well with dense and possibly streamlined objects that can descend through the rising oil. As they accumulate, those objects couple to the oil via viscous and drag forces and increase the oil's effective density. When its effective density exceeds that of the earth's crust, the oil will have essentially stopped flowing.

Bloomfield, Louis A

2010-01-01T23:59:59.000Z

150

Hanford Single-Shell Tank Leak Causes and Locations - 241-A Farm  

Science Conference Proceedings (OSTI)

This document identifies 241-A Tank Farm (A Farm) leak causes and locations for the 100 series leaking tanks (241-A-104 and 241-A-105) identified in RPP-ENV-37956, Hanford A and AX Farm Leak Assessment Report. This document satisfies the A Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

Girardot, Crystal L.; Harlow, Donald G.

2013-09-10T23:59:59.000Z

151

Hanford Single-Shell Tank Leak Causes and Locations - 241-C Farm  

SciTech Connect

This document identifies 241-C Tank Farm (C Farm) leak causes and locations for the 100 series leaking tanks (241-C-101 and 241-C-105) identified in RPP-RPT-33418, Rev. 2, Hanford C-Farm Leak Inventory Assessments Report. This document satisfies the C Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

Girardot, Crystal L.; Harlow, Donald G.

2013-07-30T23:59:59.000Z

152

Hanford Single-Shell Tank Leak Causes and Locations - 241-U Farm  

SciTech Connect

This document identifies 241-U Tank Farm (U Farm) leak causes and locations for the 100 series leaking tanks (241-U-104, 241-U-110, and 241-U-112) identified in RPP-RPT-50097, Rev. 0, Hanford 241-U Farm Leak Inventory Assessment Report. This document satisfies the U-Farm portion of the target (T04) in the Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

Girardot, Crystal L.; Harlow, Donald G.

2013-12-02T23:59:59.000Z

153

Intelligent Coatings for Location And Detection of Leaks (IntelliCLAD...  

NLE Websites -- All DOE Office Websites (Extended Search)

alerting people of the impending danger of a gas leak. Ever since the tragic natural gas explosion of 1937 in a New London, Texas school building, various governments...

154

Hanford Single-Shell Tank Leak Causes and Locations - 241-BY and 241-TY Farm  

SciTech Connect

This document identifies 241-BY Tank Farm (BY Farm) and 241-TY Tank Farm (TY Farm) leak causes and locations for the 100 series leaking tanks (241-BY-103, 241-TY-103, 241-TY-104, 241-TY-105, and 241-TY-106) identified in RPP-RPT-43704, Hanford BY Farm Leak Assessments Report, and in RPP-RPT-42296, Hanford TY Farm Leak Assessments Report. This document satisfies the BY and TY Farm portion of the target (T04) in Hanford Federal Facility Agreement and Consent Order milestone M-045-91F.

Girardot, Crystal L.; Harlow, Donald G.

2013-11-19T23:59:59.000Z

155

Ultra high vacuum pumping system and high sensitivity helium leak detector  

DOE Patents (OSTI)

An improved helium leak detection method and apparatus are disclosed which increase the leak detection sensitivity to 10{sup {minus}13} atm cc/s. The leak detection sensitivity is improved over conventional leak detectors by completely eliminating the use of o-rings, equipping the system with oil-free pumping systems, and by introducing measured flows of nitrogen at the entrances of both the turbo pump and backing pump to keep the system free of helium background. The addition of dry nitrogen flows to the system reduces back streaming of atmospheric helium through the pumping system as a result of the limited compression ratios of the pumps for helium. 2 figs.

Myneni, G.R.

1997-12-30T23:59:59.000Z

156

Ultra high vacuum pumping system and high sensitivity helium leak detector  

DOE Patents (OSTI)

An improved helium leak detection method and apparatus are disclosed which increase the leak detection sensitivity to 10.sup.-13 atm cc s.sup.-1. The leak detection sensitivity is improved over conventional leak detectors by completely eliminating the use of o-rings, equipping the system with oil-free pumping systems, and by introducing measured flows of nitrogen at the entrances of both the turbo pump and backing pump to keep the system free of helium background. The addition of dry nitrogen flows to the system reduces backstreaming of atmospheric helium through the pumping system as a result of the limited compression ratios of the pumps for helium.

Myneni, Ganapati Rao (Yorktown, VA)

1997-01-01T23:59:59.000Z

157

Method of locating a leaking fuel element in a fast breeder power reactor  

DOE Patents (OSTI)

Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

Honekamp, John R. (Downers Grove, IL); Fryer, Richard M. (Idaho Falls, ID)

1978-01-01T23:59:59.000Z

158

Estimation of Gas Leak Rates Through Very Small Orifices  

Office of Scientific and Technical Information (OSTI)

Estimation of Gas Leak Rates Estimation of Gas Leak Rates Through Very Small Orifices and Channels by Herbert J. Bomelburg February 1977 Prepared for the Nuclear Regulatory Commission -..- Pacific Northwest Laboratories Th% report was preparrd is an accceullt r.84 work spoi.wr~d by the Un~ted States Governmect. Kettker t > ~ United States nor the L'nited states 'rl:clczr 1tcgl;l;:cry Cornmiszion. :or ally c! their e m p i o y e ~ , nor any of chcrr contractors, subcontraao~r, a . tlveir rrn~invct?t-, r.~aies any H r r l a tty, cxpreji o r implied, or ?.;+~nics any !egA liability or rcrpocsibility for iirc accuracy. zcm:lc.~cn~ss 01 ~rscf.~!ccss -,f an). i?fzrxat-on, 3Poar.i:b4. prodiict cr I.m)cess disclosed, or repreen:.; :hi.: i;s i43? wott:rl n.;\ irlfringe pr ivzrc:i*l u w x o :ig.~ts.

159

You won`t find these leaks with a blower door: The latest in {open_quotes}leaking electricity{close_quotes} in homes  

SciTech Connect

Leaking electricity is the energy consumed by appliances when they are switched off or not performing their principal functions. Field measurements in Florida, California, and Japan show that leaking electricity represents 50 to 100 Watts in typical homes, corresponding to about 5 GW of total electricity demand in the United States. There are three strategies to reduce leaking electricity: eliminate leakage entirely, eliminate constant leakage and replace with intermittent charge plus storage, and improve efficiency of conversion. These options are constrained by the low value of energy savings-less than $5 per saved Watt. Some technical and lifestyle solutions are proposed. 13 refs., 1 fig., 2 tabs.

Rainer, L. [Davis Energy Group, CA (United States); Greenberg, S.; Meier, A. [Lawrence Berkeley National Lab., CA (United States)

1996-08-01T23:59:59.000Z

160

A CANDU Severe Accident Analysis  

Science Conference Proceedings (OSTI)

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

Negut, Gheorghe; Catana, Alexandru [Institute for Nuclear Research, 1, Compului Str., Mioveni, PO Box 78, 0300 Pitesti (Romania); Prisecaru, Ilie [University Politehnica Bucharest (Romania)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

A knowledge-based decision support system for shipboard damage control  

Science Conference Proceedings (OSTI)

The operational complexity of modern ships requires the use of advanced applications, called damage control systems (DCSs), able to assist crew members in the effective handling of dangerous events and accidents. In this article we describe the development ... Keywords: Damage control system, Decision support system, Expert system, Kill card, Knowledge-based system, Shipboard management

F. Calabrese; A. Corallo; A. Margherita; A. A. Zizzari

2012-07-01T23:59:59.000Z

162

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 2, Part 1C: Analysis of core damage frequency from internal events for plant operational State 5 during a refueling outage, Main report (Sections 11--14)  

SciTech Connect

This document contains the accident sequence analysis of internally initiated events for Grand Gulf, Unit 1 as it operates in the Low Power and Shutdown Plant Operational State 5 during a refueling outage. The report documents the methodology used during the analysis, describes the results from the application of the methodology, and compares the results with the results from two full power analyses performed on Grand Gulf.

Whitehead, D. [Sandia National Labs., Albuquerque, NM (United States); Darby, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States); Yakle, J. [Science Applications International Corp., Albuquerque, NM (United States)] [and others

1994-06-01T23:59:59.000Z

163

Naval Spent Fuel Rail Shipment Accident Exercise Objectives ...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives...

164

Severe Accident Management Guidance Technical Basis Report: Volumes 1 and 2  

Science Conference Proceedings (OSTI)

Severe accident management guidance encompasses actions that would be taken to recover from a damaged core condition and to prevent or mitigate the release of fission products. This report provides the technical basis for developing such guidance by the nuclear steam supply system owners groups.

1993-04-01T23:59:59.000Z

165

Markov Model of Severe Accident Progression and Management  

SciTech Connect

The earthquake and tsunami that hit the nuclear power plants at the Fukushima Daiichi site in March 2011 led to extensive fuel damage, including possible fuel melting, slumping, and relocation at the affected reactors. A so-called feed-and-bleed mode of reactor cooling was initially established to remove decay heat. The plan was to eventually switch over to a recirculation cooling system. Failure of feed and bleed was a possibility during the interim period. Furthermore, even if recirculation was established, there was a possibility of its subsequent failure. Decay heat has to be sufficiently removed to prevent further core degradation. To understand the possible evolution of the accident conditions and to have a tool for potential future hypothetical evaluations of accidents at other nuclear facilities, a Markov model of the state of the reactors was constructed in the immediate aftermath of the accident and was executed under different assumptions of potential future challenges. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accident. The work began in mid-March and continued until mid-May 2011. The analysis had the following goals: (1) To provide an overall framework for describing possible future states of the damaged reactors; (2) To permit an impact analysis of 'what-if' scenarios that could lead to more severe outcomes; (3) To determine approximate probabilities of alternative end-states under various assumptions about failure and repair times of cooling systems; (4) To infer the reliability requirements of closed loop cooling systems needed to achieve stable core end-states and (5) To establish the importance for the results of the various cooling system and physical phenomenological parameters via sensitivity calculations.

Bari, R.A.; Cheng, L.; Cuadra,A.; Ginsberg,T.; Lehner,J.; Martinez-Guridi,G.; Mubayi,V.; Pratt,W.T.; Yue, M.

2012-06-25T23:59:59.000Z

166

Model falsification diagnosis and sensor placement for leak detection in pressurized pipe networks  

Science Conference Proceedings (OSTI)

Pressurized pipe networks used for fresh-water distribution can take advantage of recent advances in sensing technologies and data-interpretation to evaluate their performance. In this paper, a leak-detection and a sensor placement methodology are proposed ... Keywords: Data interpretation, Leak detection, Sensor placement, System identification, Water distribution

James-A. Goulet, Sylvain Coutu, Ian F. C. Smith

2013-04-01T23:59:59.000Z

167

An Evaluation of Time Dependent Leak Rates in Degraded Steam Generator Tubing  

Science Conference Proceedings (OSTI)

Argonne National Laboratory (ANL) has performed leak rate testing of degraded steam generator tubing for a number of years as part of the Steam Generator Tube Integrity Program, under the sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. This document describes the results of a review and evaluation of ANL time-dependent leak rate information.

2007-12-13T23:59:59.000Z

168

Early Tube Leak Detection in a HRSG Application Using Acoustic Monitoring Technology  

Science Conference Proceedings (OSTI)

Acoustic monitoring has become an essential part of early tube leak detection for conventional boilers. Acoustic monitoring is intended for steam leak detection in pressurized vessels, including power boilers, recovery boilers, and feedwater heaters. The system performs acoustic monitoring by continuously measuring the internal sounds from the boiler, signaling an alarm when the sound exceeds a preset threshold for a ...

2012-12-12T23:59:59.000Z

169

Purify: Fast detection of memory leaks and access errors  

E-Print Network (OSTI)

This paper describes Purifyru, a software testing and quality assurance Ool that detects memory leaks and access erors. Purify inserts additional checking instructions directly into the object code produced by existing compilers. These instructions check every memory read and write performed by the program-under-test and detect several types of access errors, such as reading uninitialized memory or witing to freed memory. Purify inserts checking logic into all of the code in a program, including third-party and vendor object-code libraries, and verifies system call interfaces. In addition, Purify tracks memory usage and identifies individual memory leals using a novel adaptation of garbage collection techniques. Purify produce standard executable files compatible with existing debuggers, and currently runs on Sun Microsystems ' SPARC family of workstations. Purify's neafly-comprehensive memory access checking slows the target program down typically by less than a facor of three and has resulted in significantly more reliable software for several development goups. L.

Reed Hastings; Bob Joyce

1991-01-01T23:59:59.000Z

170

Fuel leak detection apparatus for gas cooled nuclear reactors  

SciTech Connect

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

Burnette, Richard D. (San Diego, CA)

1977-01-01T23:59:59.000Z

171

Leak rate analysis of the Westinghouse Reactor Coolant Pump  

SciTech Connect

An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs.

Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

1985-07-01T23:59:59.000Z

172

Methods for Integrated Leak Detection Inference at CO2 Sequestration Sites  

NLE Websites -- All DOE Office Websites (Extended Search)

Methods for Integrated Leak Detection Inference at CO2 Sequestration Sites Methods for Integrated Leak Detection Inference at CO2 Sequestration Sites Speaker(s): Mitchell Small Date: March 23, 2010 - 12:00pm Location: 90-3122 This seminar will explain a methodology for combining site characterization and soil CO2 monitoring for detecting leaks at geologic CO2 sequestration sites. Near surface CO2 fluxes resulting from a leak are simulated using the TOUGH2 model for different values of soil permeability, leakage rate and vadose zone thickness. Natural background soil CO2 flux rates are characterized by a Bayesian hierarchical model that predicts the background flux as a function of soil temperature. A presumptive leak is assumed if the monitored flux rate exceeds a critical value corresponding to a very high (e.g., 99%) prediction interval for the natural flux conditioned on

173

NETL: News Release - Field Testing Underway of Remote Sensor Gas Leak  

NLE Websites -- All DOE Office Websites (Extended Search)

September 16, 2004 September 16, 2004 Field Testing Underway of Remote Sensor Gas Leak Detection Systems CASPER, WY-An extensive field test that will document and demonstrate how effective technologies are in remotely detecting natural gas leaks is being held September 13-17, as the Department of Energy simulates natural gas leaks along a predetermined course at DOE's Rocky Mountain Oilfield Testing Center (RMOTC). Low-flying aircraft, satellites and special ground vehicles carrying advanced leak detection sensors will participate as representatives of the gas industry and potential technology manufacturers observe the technologies in a real-world environment and evaluate their readiness for commercialization. The test plan was devised with strong input from an industry advisory board and test participants to compare the effectiveness of several gas-leak detection devices from ground, air and satellite based platforms.

174

NETL: News Release - Vehicle-Mounted Natural Gas Leak Detector Passes Key  

NLE Websites -- All DOE Office Websites (Extended Search)

October 2, 2003 October 2, 2003 Vehicle-Mounted Natural Gas Leak Detector Passes Key "Road Test" Spots Natural Gas Leaks from 30 Feet Away At Speeds Approaching 20 Miles Per Hour Handheld Prototype Gas Detector Now Being Outfitted as a Van-Mounted Unit PSI has modified this early prototype of a handheld remote natural gas detector to operate from a moving vehicle. ANDOVER, MA - Physical Sciences Inc. (PSI) recently conducted a successful test of its mobile natural gas detector at the company's research facilities in Andover, Mass. PSI's prototype leak detector demonstrated its ability to spot natural gas leaks from a distance of up to 30 feet from a vehicle moving at speeds approaching 20 miles per hour. In the United States, significant resources are devoted annually to leak

175

Hanford Determines Double-Shell Tank Leaked Waste From Inner Tank |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Determines Double-Shell Tank Leaked Waste From Inner Tank Determines Double-Shell Tank Leaked Waste From Inner Tank Hanford Determines Double-Shell Tank Leaked Waste From Inner Tank October 22, 2012 - 12:00pm Addthis Media Contacts Lori Gamache, ORP 509-372-9130 John Britton, WRPS 509-376-5561 RICHLAND - The Department of Energy's Office of River Protection (ORP), working with its Hanford tank operations contractor Washington River Protection Solutions, has determined that there is a slow leak of chemical and radioactive waste into the annulus space in Tank AY-102, the approximately 30-inch area between the inner primary tank and the outer tank that serves as the secondary containment for these types of tanks. This is the first time a double-shell tank (DST) leak from the primary tank into the annulus has been identified. There is no indication of waste in

176

Evaluation and refinement of leak-rate estimation models. Revision 1  

Science Conference Proceedings (OSTI)

Leak-rate estimation models are important elements in developing a leak-beforebreak methodology in piping integrity and safety analyses. Existing thermalhydraulic and crack-opening-area models used in current leak-rate estimations have been incorporated into a single computer code for leak-rate estimation. The code is called SQUIRT, which stands for Seepage Quantification of Upsets In Reactor Tubes. The SQUIRT program has been validated by comparing its thermalhydraulic predictions with the limited experimental data that have been published on two-phase flow through slits and cracks, and by comparing its crack-opening-area predictions with data from the Degraded Piping Program. In addition, leak-rate experiments were conducted to obtain validation data for a circumferential fatigue crack in a carbon steel pipe girth weld.

Paul, D.D.; Ahmad, J.; Scott, P.M.; Flanigan, L.F.; Wilkowski, G.M. [Battelle, Columbus, OH (United States)

1994-06-01T23:59:59.000Z

177

Accident management for indian pressurized heavy water reactors  

Science Conference Proceedings (OSTI)

Indian nuclear power program as of now is mainly based on Pressurized Heavy Water Reactors (PHWRs). Operating Procedures for normal power operation and Emergency Operating Procedures for operational transients and accidents within design basis exist for all Indian PHWRs. In addition, on-site and off-site emergency response procedures are also available for these NPPs. The guidelines needed for severe accidents mitigation are now formally being documented for Indian PHWRs. Also, in line with International trend of having symptom based emergency handling, the work is in advanced stage for preparation of symptom-based emergency operating procedures. Following a plant upset condition; a number of alarms distributed in different information systems appear in the control room to aid operator to identify the nature of the event. After identifying the event, appropriate intervention in the form of event based emergency operating procedure is put into use by the operating staff. However, if the initiating event cannot be unambiguously identified or after the initial event some other failures take place, then the selected event based emergency operating procedure will not be optimal. In such a case, reactor safety is ensured by monitoring safety functions (depicted by selected plant parameters grouped together) throughout the event handling so that the barriers to radioactivity release namely, fuel and fuel cladding, primary heat transport system integrity and containment remain intact. Simultaneous monitoring of all these safety functions is proposed through status trees and this concept will be implemented through a computer-based system. For beyond design basis accidents, event sequences are identified which may lead to severe core damage. As part of this project, severe accident mitigation guidelines are being finalized for the selected event sequences. The paper brings out the details of work being carried out for Indian PHWRs for symptom based event handling and severe accident management. (authors)

Hajela, S.; Grover, R.; Ghadge, S.G.; Bajaj, S.S. [Directorate of Safety, Nuclear Power Corporation of India Limited Nabhikiya Urja Bhawan, Anushakti Nagar, Mumbai-400 094 (India)

2006-07-01T23:59:59.000Z

178

Chernobyl Nuclear Accident | National Nuclear Security Administration  

NLE Websites -- All DOE Office Websites (Extended Search)

Chernobyl Nuclear Accident | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

179

Evolvable neural networks ensembles for accidents diagnosis  

Science Conference Proceedings (OSTI)

Prediction and diagnosis of nuclear accidents is one of the most important tasks for nuclear safety. Since accurate diagnosis of nuclear accident is a very important issue for avoidance of disastrous outcomes, it is more desirable to make a decision ... Keywords: ensembles, neuroevolution, nuclear accidents

Hany Sallam; Carlo S. Regazzoni; Ihab Talkhan; Amir Atiya

2008-07-01T23:59:59.000Z

180

Thermal Imaging of Canals for Remote Detection of Leaks: Evaluation in the United Irrigation District  

E-Print Network (OSTI)

This report summarizes our initial analysis of the potential of thermal imaging for detecting leaking canals and pipelines. Thermal imagery (video format) was obtained during a fly over of a portion of the main canal of United Irrigation District. The video was processed to produce individual images, and 45 potential sites were identified as having possible canal leakage problems (see Appendix I for all 45 thermal images). District Management System Team personnel traveled to 11 of the 45 sites to determine if canal leakage was actually occurring. Of the 11 sites, 10 had leakage problems. Thus, thermal image analysis had a success rate of 91% for leak detection. Two sites had leaks classified as severe by the DMS Team. This report also provides a detailed analysis of 4 sites, 3 with leaks and 1 without. For each site, photographs are included showing the source of the leak and/or condition of the canal segment. A literature review of thermal imagery for leak detection is included in Appendix II. Our findings and recommendations are as following: 1. thermal imaging is a promising technique for evaluation of canal conditions and leak detection; 2. the district provide should provide personnel to help the DMS Team verify the remaining 34 sites; and 3. the district should consider correcting the problems identified at sites 7 and 8.

Huang, Yanbo; Fipps, Guy

2008-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Retrospection of Chernobyl nuclear accident for decision analysis concerning remedial actions in Ukraine  

SciTech Connect

It is considered the efficacy of decisions concerning remedial actions when of-site radiological monitoring in the early and (or) in the intermediate phases was absent or was not informative. There are examples of such situations in the former Soviet Union where many people have been exposed: releases of radioactive materials from 'Krasnoyarsk-26' into Enisey River, releases of radioactive materials from 'Chelabinsk-65' (the Kishtim accident), nuclear tests at the Semipalatinsk Test Site, the Chernobyl nuclear accident etc. If monitoring in the early and (or) in the intermediate phases is absent the decisions concerning remedial actions are usually developed on the base of permanent monitoring. However decisions of this kind may be essentially erroneous. For these cases it is proposed to make retrospection of radiological data of the early and intermediate phases of nuclear accident and to project decisions concerning remedial actions on the base of both retrospective data and permanent monitoring data. In this Report the indicated problem is considered by the example of the Chernobyl accident for Ukraine. Their of-site radiological monitoring in the early and intermediate phases was unsatisfactory. In particular, the pasture-cow-milk monitoring had not been made. All official decisions concerning dose estimations had been made on the base of measurements of {sup 137}Cs in body (40 measurements in 135 days and 55 measurements in 229 days after the Chernobyl accident). For the retrospection of radiological data of the Chernobyl accident dynamic model has been developed. This model has structure similar to the structure of Pathway model and Farmland model. Parameters of the developed model have been identified for agricultural conditions of Russia and Ukraine. By means of this model dynamics of 20 radionuclides in pathways and dynamics of doses have been estimated for the early, intermediate and late phases of the Chernobyl accident. The main results are following: - During the first year after the Chernobyl accident 75-93% of Commitment Effective Dose had been formed; - During the first year after the Chernobyl accident 85-90% of damage from radiation exposure had been formed. During the next 50 years (the late phase of accident) only 10-15% of damage from radiation exposure will have been formed; - Remedial actions (agricultural remedial actions as most effective) in Ukraine are intended for reduction of the damage from consumption of production which is contaminated in the late phase of accident. I.e. agricultural remedial actions have been intended for minimization only 10 % of the total damage from radiation exposure; - Medical countermeasures can minimize radiation exposure damage by an order of magnitude greater than agricultural countermeasures. - Thus, retrospection of nuclear accident has essentially changed type of remedial actions and has given a chance to increase effectiveness of spending by an order of magnitude. This example illustrates that in order to optimize remedial actions it is required to use data of retrospection of nuclear accidents in all cases when monitoring in the early and (or) intermediate phases is unsatisfactory. (author)

Georgievskiy, Vladimir [Russian Research Center 'Kurchatov Insitute', Kurchatov Sq., 1, 123182 Moscow (Russian Federation)

2007-07-01T23:59:59.000Z

182

USING AN ADAPTER TO PERFORM THE CHALFANT-STYLE CONTAINMENT VESSEL PERIODIC MAINTENANCE LEAK RATE TEST  

Science Conference Proceedings (OSTI)

Recently the Packaging Technology and Pressurized Systems (PT&PS) organization at the Savannah River National Laboratory was asked to develop an adapter for performing the leak-rate test of a Chalfant-style containment vessel. The PT&PS organization collaborated with designers at the Department of Energy's Pantex Plant to develop the adapter currently in use for performing the leak-rate testing on the containment vessels. This paper will give the history of leak-rate testing of the Chalfant-style containment vessels, discuss the design concept for the adapter, give an overview of the design, and will present results of the testing done using the adapter.

Loftin, B.; Abramczyk, G.; Trapp, D.

2011-06-03T23:59:59.000Z

183

Demonstration of rapid and sensitive module leak certification for space station freedom  

SciTech Connect

A leak detection and quantification demonstration using perflurocarbon tracer (PFT) technology was successfully performed at the NASA Marshall Space Flight Center on January 25, 1991. The real-time Dual Trap Analyzer (DTA) at one-half hour after the start of the first run gave an estimated leak rate of 0.7 mL/min. This has since been refined to be 1.15 {plus minus} 0.09 mL/min. The leak rates in the next three runs were determined to be 9.8 {plus minus} 0.7, {minus}0.4 {plus minus} 0.3, and 76 {plus minus} 6 mL/min, respectively. The theory on leak quantification in the steady-state and time-dependent modes for a single zone test facility was developed and applied to the above determinations. The laboratory PFT analysis system gave a limit-of-detection (LOD) of 0.05 fL for ocPDCH. This is the tracer of choice and is about 100-fold better than that for the DTA. Applied to leak certification, the LOD is about 0.00002 mL/s (0.000075 L/h), a 5 order-of-magnitude improvement over the original leak certification specification. Furthermore, this limit can be attained in a measurement period of 3 to 4 hours instead of days, weeks, or months. A new Leak Certification Facility is also proposed to provide for zonal (three zones) determination of leak rates. The appropriate multizone equations, their solutions, and error analysis have already been derived. A new concept of seal-integrity certification has been demonstrated for a variety of controlled leaks in the range of module leak testing. High structural integrity leaks were shown to have a linear dependence of flow on {Delta}p. The rapid determination of leak rates at different pressures is proposed and is to be determined while subjecting the module to other external force-generating parameters such as vibration, torque, solar intensity, etc. 13 refs.

Dietz, R.N.; Goodrich, R.W. (Brookhaven National Lab., Upton, NY (United States))

1991-03-01T23:59:59.000Z

184

Demonstration of rapid and sensitive module leak certification for space station freedom. Final report  

SciTech Connect

A leak detection and quantification demonstration using perflurocarbon tracer (PFT) technology was successfully performed at the NASA Marshall Space Flight Center on January 25, 1991. The real-time Dual Trap Analyzer (DTA) at one-half hour after the start of the first run gave an estimated leak rate of 0.7 mL/min. This has since been refined to be 1.15 {plus_minus} 0.09 mL/min. The leak rates in the next three runs were determined to be 9.8 {plus_minus} 0.7, {minus}0.4 {plus_minus} 0.3, and 76 {plus_minus} 6 mL/min, respectively. The theory on leak quantification in the steady-state and time-dependent modes for a single zone test facility was developed and applied to the above determinations. The laboratory PFT analysis system gave a limit-of-detection (LOD) of 0.05 fL for ocPDCH. This is the tracer of choice and is about 100-fold better than that for the DTA. Applied to leak certification, the LOD is about 0.00002 mL/s (0.000075 L/h), a 5 order-of-magnitude improvement over the original leak certification specification. Furthermore, this limit can be attained in a measurement period of 3 to 4 hours instead of days, weeks, or months. A new Leak Certification Facility is also proposed to provide for zonal (three zones) determination of leak rates. The appropriate multizone equations, their solutions, and error analysis have already been derived. A new concept of seal-integrity certification has been demonstrated for a variety of controlled leaks in the range of module leak testing. High structural integrity leaks were shown to have a linear dependence of flow on {Delta}p. The rapid determination of leak rates at different pressures is proposed and is to be determined while subjecting the module to other external force-generating parameters such as vibration, torque, solar intensity, etc. 13 refs.

Dietz, R.N.; Goodrich, R.W. [Brookhaven National Lab., Upton, NY (United States)

1991-03-01T23:59:59.000Z

185

Plan for support of large-plant (post-CRBR) needs in large-leak sodium-water reaction area  

SciTech Connect

Work in the large leak test and analysis area of steam generator development has been carried out at GE-ARSD under 189a SG037 since 1973. The currently planned master schedule for the SG037 program is shown. Principal activities are the large leak testing program being carried out at the Large Leak Test Rig and the analysis methods development. The plan for supporting the large plant (post-CRBR) needs in the large leak sodium-water reaction area is outlined. Most of the needs will be answered in the current SG037 large leak program. (DLC)

Whipple, J.C.

1980-03-01T23:59:59.000Z

186

Schools - CPU Damage  

Science Conference Proceedings (OSTI)

This power quality (PQ) case study presents the investigation of computers in a school computer lab that were being damaged.

2003-12-31T23:59:59.000Z

187

Corrosion and Hydrogen Damage  

Science Conference Proceedings (OSTI)

Mar 5, 2013 ... Advanced Materials and Reservoir Engineering for Extreme Oil & Gas Environments: Corrosion and Hydrogen Damage Sponsored by: TMS...

188

An evaluation of the leakage potential of a personnel airlock subject to severe accident loads  

SciTech Connect

A systematic investigation of the performance of light water reactor containment buildings subject to severe accident loads must include the consideration of leakage between the sealing surfaces of penetrations. As part of its work on containment integrity for the US Nuclear Regulatory Commission (USNRC), Sandia National Laboratories is developing test validated methods for predicting leakage from mechanical penetrations. The primary emphasis has been on large diameter operable penetrations, such as equipment hatches, personnel airlocks, and drywell heads. Several studies conducted for the USNRC have identified leakage from personnel airlocks as a potentially significant failure mode of containment buildings subject to severe accident loads, including Barnes et al. (1984) and Shackelford et al. (1985). Barnes et al (1984) conducted finite element analyses to predict separation of the sealing surfaces on the door and bulkhead, but they did not consider elevated temperature effects and they did not take credit for the performance of the seal material in calculating leak areas. To the author's knowledge, personnel airlock designs with flat bulkhead/door assemblies and seals have never been tested under severe accident conditions, i.e., elevated temperatures and pressure. This paper will describe preliminary analyses and plans for testing a full-size personnel airlock.

Clauss, D.B.

1987-01-01T23:59:59.000Z

189

Markov Model of Accident Progression at Fukushima Daiichi  

DOE Green Energy (OSTI)

On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.

Cuadra A.; Bari R.; Cheng, L-Y; Ginsberg, T.; Lehner, J.; Martinez-Guridi, G.; Mubayi, V.; Pratt, T.; Yue, M.

2012-11-11T23:59:59.000Z

190

BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Field Sites Power Marketing Administration Other Agencies You are here Home BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010 (3 of 4) BP Oil Spill Footage...

191

BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Field Sites Power Marketing Administration Other Agencies You are here Home BP Oil Spill Footage (High Def) - Leak at 4850' - June 3 2010 (2 of 4) BP Oil Spill Footage...

192

BP Oil Spill Footage (High Def) - Leak at 4840' - June 3 2010...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Field Sites Power Marketing Administration Other Agencies You are here Home BP Oil Spill Footage (High Def) - Leak at 4840' - June 3 2010 (1 of 4) BP Oil Spill Footage...

193

Bhopal Gas Leak: A Numerical Investigation of the Prevailing Meteorological Conditions  

Science Conference Proceedings (OSTI)

A three-dimensional mesoscale model was used to understand the meteorological conditions and the influence of the terrain on the local flow pattern during the Bhopal methyl isocyanate (MIC) gas leak. The study reveals that under the prevailing ...

Maithili Sharan; S. G. Gopalakrishnan; R. T. McNider; M. P. Singh

1996-10-01T23:59:59.000Z

194

Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability  

DOE Patents (OSTI)

A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

Hunsbedt, A.; Boardman, C.E.

1995-04-11T23:59:59.000Z

195

Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability  

SciTech Connect

A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1995-01-01T23:59:59.000Z

196

AIRBORNE, OPTICAL REMOTE SENSNG OF METHANE AND ETHANE FOR NATURAL GAS PIPELINE LEAK DETECTION  

SciTech Connect

Ophir Corporation was awarded a contract by the U. S. Department of Energy, National Energy Technology Laboratory under the Project Title ''Airborne, Optical Remote Sensing of Methane and Ethane for Natural Gas Pipeline Leak Detection'' on October 14, 2002. The scope of the work involved designing and developing an airborne, optical remote sensor capable of sensing methane and, if possible, ethane for the detection of natural gas pipeline leaks. Flight testing using a custom dual wavelength, high power fiber amplifier was initiated in February 2005. Ophir successfully demonstrated the airborne system, showing that it was capable of discerning small amounts of methane from a simulated pipeline leak. Leak rates as low as 150 standard cubic feet per hour (scf/h) were detected by the airborne sensor.

Jerry Myers

2005-04-15T23:59:59.000Z

197

Pre-test evaluation of LLTR Series II Test A-6. [Large Leak Test Facility  

SciTech Connect

Purpose of this report is to present pre-test predictions of pressure histories for the A6 test to be conducted in the Large Leak Test Facility (LLTF) at the Energy Technology Engineering Center. A6 is part of a test program being conducted to evaluate the effects of leaks produced by a double-ended guillotine rupture of a single tube. A6 will provide data on the CRBR prototypical double rupture disc performance.

Knittle, D.

1980-11-01T23:59:59.000Z

198

An Evaluation of Time Dependent Leak Rates in Degraded Steam Generator Tubing  

Science Conference Proceedings (OSTI)

The U.S. Nuclear Regulatory Commission sponsored leak rate tests of steam generator (SG) tubing with stress corrosion cracks and electrodischarged machining (EDM) notches at Argonne National Laboratory (ANL). Some test specimens displayed time-dependent leak rate increases when the pressure was held constant. Post-test visual examination clearly revealed that the outside diameter (OD) crack length of these specimens had increased. It was suspected that fatigue due to jet/structure interaction was respons...

2008-10-08T23:59:59.000Z

199

Leak Detection and H2 Sensor Development for Hydrogen Applications  

DOE Green Energy (OSTI)

The objectives of this report are: (1) Develop a low cost, low power, durable, and reliable hydrogen safety sensor for a wide range of vehicle and infrastructure applications; (2) Continually advance test prototypes guided by materials selection, sensor design, electrochemical R&D investigation, fabrication, and rigorous life testing; (3) Disseminate packaged sensor prototypes and control systems to DOE Laboratories and commercial parties interested in testing and fielding advanced prototypes for cross-validation; (4) Evaluate manufacturing approaches for commercialization; and (5) Engage an industrial partner and execute technology transfer. Recent developments in the search for sustainable and renewable energy coupled with the advancements in fuel cell powered vehicles (FCVs) have augmented the demand for hydrogen safety sensors. There are several sensor technologies that have been developed to detect hydrogen, including deployed systems to detect leaks in manned space systems and hydrogen safety sensors for laboratory and industrial usage. Among the several sensing methods electrochemical devices that utilize high temperature-based ceramic electrolytes are largely unaffected by changes in humidity and are more resilient to electrode or electrolyte poisoning. The desired sensing technique should meet a detection threshold of 1% (10,000 ppm) H{sub 2} and response time of {approx_equal}1 min, which is a target for infrastructure and vehicular uses. Further, a review of electrochemical hydrogen sensors by Korotcenkov et.al and the report by Glass et.al suggest the need for inexpensive, low power, and compact sensors with long-term stability, minimal cross-sensitivity, and fast response. This view has been largely validated and supported by the fuel cell and hydrogen infrastructure industries by the NREL/DOE Hydrogen Sensor Workshop held on June 8, 2011. Many of the issues preventing widespread adoption of best-available hydrogen sensing technologies available today outside of cost, derive from excessive false positives and false negatives arising from signal drift and unstable sensor baseline; both of these problems necessitate the need for unacceptable frequent calibration.

Brosha, Eric L. [Los Alamos National Laboratory

2012-07-10T23:59:59.000Z

200

Computerized Accident/Incident Reporting System  

NLE Websites -- All DOE Office Websites (Extended Search)

Accident Recordkeeping and Reporting Accident Recordkeeping and Reporting Accident/Incident Recordkeeping and Reporting CAIRS logo Computerized Accident Incident Reporting System CAIRS Database The Computerized Accident/Incident Reporting System is a database used to collect and analyze DOE and DOE contractor reports of injuries, illnesses, and other accidents that occur during DOE operations. Injury and Illness Dashboard The Dashboard provides an alternate interface to CAIRS information. The initial release of the Dashboard allows analysis of composite DOE-wide information and summary information by Program Office, and site. Additional data feature are under development. CAIRS Registration Form CAIRS is a Government computer system and, as such, has security requirements that must be followed. Access to the

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Recommendations for Analyzing Accidents Under NEPA  

Energy.gov (U.S. Department of Energy (DOE))

This DOE guidance clarifies and supplements "Recommendations for the Preparation of Environmental Assessments and Environmental Impact Statements." It focuses on principles of accident analyses under NEPA.

202

Accident Tolerant Fuels for Light Water Reactors  

Science Conference Proceedings (OSTI)

Presentation Title, Accident Tolerant Fuels for Light Water Reactors. Author(s), Steven J. Zinkle, Kurt A. Terrani, Lance L. Snead. On-Site Speaker (Planned)...

203

Systematics of Reconstructed Process Facility Criticality Accidents  

SciTech Connect

The systematics of the characteristics of twenty-one criticality accidents occurring in nuclear processing facilities of the Russian Federation, the United States, and the United Kingdom are examined. By systematics the authors mean the degree of consistency or agreement between the factual parameters reported for the accidents and the experimentally known conditions for criticality. The twenty-one reported process criticality accidents are not sufficiently well described to justify attempting detailed neutronic modeling. However, results of classic hand calculations confirm the credibility of the reported accident conditions.

Pruvost, N.L.; McLaughlin, T.P.; Monahan, S.P.

1999-09-19T23:59:59.000Z

204

ORISE: REAC/TS Radiation Accident Registries  

NLE Websites -- All DOE Office Websites (Extended Search)

Accident Registries The Radiation Emergency Assistance CenterTraining Site (REACTS) at the Oak Ridge Institute for Science and Education (ORISE) maintains a number of radiation...

205

Accident Investigation Report Plutonium Contamination in the...  

NLE Websites -- All DOE Office Websites (Extended Search)

Accident Investigation Report Plutonium Contamination in the Zero Power Physics Reactor Facility at the Idaho National Laboratory, November 8, 2011 January 2012 Disclaimer...

206

RELAP5 Application to Accident Analysis of the NIST Research Reactor  

Science Conference Proceedings (OSTI)

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

2012-03-18T23:59:59.000Z

207

Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling  

E-Print Network (OSTI)

Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and...

Saad, Hend Mohammed El Sayed; Wahab, Moustafa Aziz Abd El

2013-01-01T23:59:59.000Z

208

Estimates of the financial consequences of nuclear-power-reactor accidents  

SciTech Connect

This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents.

Strip, D.R.

1982-09-01T23:59:59.000Z

209

Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling  

E-Print Network (OSTI)

Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and satisfactory agreement is found.

Hend Mohammed El Sayed Saad; Hesham Mohammed Mohammed Mansour; Moustafa Aziz Abd El Wahab

2013-06-05T23:59:59.000Z

210

Locating of leaks in water-cooled generator stator bars using perfluorocarbon tracers  

SciTech Connect

Water cooled stator bars in power plant generators often fail during the maintenance cycle due to water leakage. After the hydrogen pressure in the generator shell has been released water can leak through cracks in the copper and through the insulation. Leaking bars, but not the leaks themselves, are detected with so-called ``hi-pot`` (high potential) tests where direct electrical current is applied to the stator bar windings. A study initiated by ConEd and Brookhaven`s Tracer Technology Center to explore the cause of these leakage problems to determine if the failures originate in the manufacturing process or are created in service by phase related torque stresses. To this purpose bars that had failed the hi-pot test were investigated first with the insulation in place and then stripped to the bare copper. The bars were pressurized with gases containing perfluorocarbon tracers and the magnitude and location of the leaks was detected by using tracers technology principles and instruments such as the ``double source`` method and the Dual Trap Analyzer. In the second part of the project the windings within a generator were tested in-situ for leaks during an outage using tracer principles. Recommendations are given suggesting the shut down of stator bar cooling water before hydrogen bleeding during outages and a revision of the current vent flow rate. The new standard should establish a reasonable leak rate for the stator bar windings proper and exclude leakage of pump seals and connections. Testing during the maintenance cycle in generators should include routine tracer leak detection following the hi-pot test.

Loss, W.M.; Dietz, R.N.

1991-09-01T23:59:59.000Z

211

Locating of leaks in water-cooled generator stator bars using perfluorocarbon tracers  

SciTech Connect

Water cooled stator bars in power plant generators often fail during the maintenance cycle due to water leakage. After the hydrogen pressure in the generator shell has been released water can leak through cracks in the copper and through the insulation. Leaking bars, but not the leaks themselves, are detected with so-called hi-pot'' (high potential) tests where direct electrical current is applied to the stator bar windings. A study initiated by ConEd and Brookhaven's Tracer Technology Center to explore the cause of these leakage problems to determine if the failures originate in the manufacturing process or are created in service by phase related torque stresses. To this purpose bars that had failed the hi-pot test were investigated first with the insulation in place and then stripped to the bare copper. The bars were pressurized with gases containing perfluorocarbon tracers and the magnitude and location of the leaks was detected by using tracers technology principles and instruments such as the double source'' method and the Dual Trap Analyzer. In the second part of the project the windings within a generator were tested in-situ for leaks during an outage using tracer principles. Recommendations are given suggesting the shut down of stator bar cooling water before hydrogen bleeding during outages and a revision of the current vent flow rate. The new standard should establish a reasonable leak rate for the stator bar windings proper and exclude leakage of pump seals and connections. Testing during the maintenance cycle in generators should include routine tracer leak detection following the hi-pot test.

Loss, W.M.; Dietz, R.N.

1991-09-01T23:59:59.000Z

212

AUTOMATED LEAK DETECTION OF BURIED TANKS USING GEOPHYSICAL METHODS AT THE HANFORD NUCLEAR SITE  

SciTech Connect

At the Hanford Nuclear Site in Washington State, the Department of Energy oversees the containment, treatment, and retrieval of liquid high-level radioactive waste. Much of the waste is stored in single-shelled tanks (SSTs) built between 1943 and 1964. Currently, the waste is being retrieved from the SSTs and transferred into newer double-shelled tanks (DSTs) for temporary storage before final treatment. Monitoring the tanks during the retrieval process is critical to identifying leaks. An electrically-based geophysics monitoring program for leak detection and monitoring (LDM) has been successfully deployed on several SSTs at the Hanford site since 2004. The monitoring program takes advantage of changes in contact resistance that will occur when conductive tank liquid leaks into the soil. During monitoring, electrical current is transmitted on a number of different electrode types (e.g., steel cased wells and surface electrodes) while voltages are measured on all other electrodes, including the tanks. Data acquisition hardware and software allow for continuous real-time monitoring of the received voltages and the leak assessment is conducted through a time-series data analysis. The specific hardware and software combination creates a highly sensitive method of leak detection, complementing existing drywell logging as a means to detect and quantify leaks. Working in an industrial environment such as the Hanford site presents many challenges for electrical monitoring: cathodic protection, grounded electrical infrastructure, lightning strikes, diurnal and seasonal temperature trends, and precipitation, all of which create a complex environment for leak detection. In this discussion we present examples of challenges and solutions to working in the tank farms of the Hanford site.

CALENDINE S; SCHOFIELD JS; LEVITT MT; FINK JB; RUCKER DF

2011-03-30T23:59:59.000Z

213

The Hartford Life and Accident Insurance  

E-Print Network (OSTI)

The Hartford Life and Accident Insurance Company Group Numbers Basic Group Term Life AD&D-677984 Life and Accident Insurance Company. (Referred to as The Hartford or Hartford.) General information industry. Europ Assist has been helping customers in times of crisis for more than 46 years. They have

214

Accident states simulation: process fluids release  

Science Conference Proceedings (OSTI)

Seveso II Directive imposes for high hazardous plants quantitative risk evaluation of the major accident. In a general context the risk is defined as product between frequency and consequences of accident state. There are five steps in quantitative risk ... Keywords: hazard, hydrogen sulphide, mathematical model, release, risk, safety system, simulation

Cornelia Croitoru; Mihai Anghel; Floarea Pop; Ioan Stefanescu; Gheorghe Titescu; Mihai Patrascu; Ervin Watzlawek; Dorin Cheresdi

2008-08-01T23:59:59.000Z

215

Does Daylight Savings Time Affect Traffic Accidents?  

E-Print Network (OSTI)

This paper studies the effect of changes in accident pattern due to Daylight Savings Time (DST). The extension of the DST in 2007 provides a natural experiment to determine whether the number of traffic accidents is affected by shifts in hours of daylight using the year as control group. Using data on traffic accidents in Texas based on crash reports provided by the Texas Transportation Institute, and a difference in differences technique, this study creates a regression model to determine how significant this factor is in affecting traffic accident patterns as observed in the data. Results show that DST has no statistically significant effect on traffic accidents of all categories including (but not limited to) highway, non-highway, and accidents, accidents with injuries and no injuries, and accidents by drivers of all age-groups. This implies that the federal governments policy of DST (and its extension) has no costs incurred by a rise in motor vehicle crashes when it gets dark early.

Deen, Sophia 1988-

2012-05-01T23:59:59.000Z

216

Flight Testing of an Advanced Airborne Natural Gas Leak Detection System  

SciTech Connect

ITT Industries Space Systems Division (Space Systems) has developed an airborne natural gas leak detection system designed to detect, image, quantify, and precisely locate leaks from natural gas transmission pipelines. This system is called the Airborne Natural Gas Emission Lidar (ANGEL) system. The ANGEL system uses a highly sensitive differential absorption Lidar technology to remotely detect pipeline leaks. The ANGEL System is operated from a fixed wing aircraft and includes automatic scanning, pointing system, and pilot guidance systems. During a pipeline inspection, the ANGEL system aircraft flies at an elevation of 1000 feet above the ground at speeds of between 100 and 150 mph. Under this contract with DOE/NETL, Space Systems was funded to integrate the ANGEL sensor into a test aircraft and conduct a series of flight tests over a variety of test targets including simulated natural gas pipeline leaks. Following early tests in upstate New York in the summer of 2004, the ANGEL system was deployed to Casper, Wyoming to participate in a set of DOE-sponsored field tests at the Rocky Mountain Oilfield Testing Center (RMOTC). At RMOTC the Space Systems team completed integration of the system and flew an operational system for the first time. The ANGEL system flew 2 missions/day for the duration for the 5-day test. Over the course of the week the ANGEL System detected leaks ranging from 100 to 5,000 scfh.

Dawn Lenz; Raymond T. Lines; Darryl Murdock; Jeffrey Owen; Steven Stearns; Michael Stoogenke

2005-10-01T23:59:59.000Z

217

The Fukushima Daiichi Accident Study Information Portal  

SciTech Connect

This paper presents a description of The Fukushima Daiichi Accident Study Information Portal. The Information Portal was created by the Idaho National Laboratory as part of joint NRC and DOE project to assess the severe accident modeling capability of the MELCOR analysis code. The Fukushima Daiichi Accident Study Information Portal was created to collect, store, retrieve and validate information and data for use in reconstructing the Fukushima Daiichi accident. In addition to supporting the MELCOR simulations, the Portal will be the main DOE repository for all data, studies and reports related to the accident at the Fukushima Daiichi nuclear power station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.

Shawn St. Germain; Curtis Smith; David Schwieder; Cherie Phelan

2012-11-01T23:59:59.000Z

218

Acceptance test report for the AN valve pit leak detection and low point drain assembly mock up test procedure  

SciTech Connect

This document describes The Performance Mock-up Test Procedure for the Valve Pit Leak Detection and Low Point Drain Assembly Performance Mock-Up Test Procedure.

EWER, K.L.

1999-07-20T23:59:59.000Z

219

Leak potential index model for use in priority ranking of underground storage tanks at formerly used defense sites. Final report  

SciTech Connect

Abandoned underground storage tanks (USTs) that have not been properly closed at formerlC used defense sites (FUDS) may present potential leaking problems, spilling their hazardous contents into nearby soils, groundwater, and well water. The leaking USTs are potential sources of contaminants generally classified as containerized hazardous, toxic, and radioactive waste (CON/HTRW). CON/IITRW includes petroleum, oil, and lubricants (POL), benzene, toluene, ethylbenzene, xylene (BTEX), and radioactive waste products. The risk to the environment and population associated with the leaking USTs depends not only on the source, but on the migration pathway factor (MPF) (i.e., the ability of the medium of transport such as soil or water-to effectively transport the contaminants to the receptor) and finally on the relative vulnerability of the potential receptor. Thus, the assessment of the relative risk begins with the calculation of the potential of the UST to leak. A method of predicting the risk of leakage of these USTs is therefore desirable. presently, however, leak prediction index (LPI) models (which are used to predict the age at which a UST will leak or the probability of a UST leak at any given age) require soil data that are not readily available, or not easily and economically obtained by LPI.model users. The Warren Rogers leak prediction model was developed circa 1981, and has been used for USTs and incorporated into leak prediction models for other types of underground steel structures.

Stephenson, L.D.

1998-03-01T23:59:59.000Z

220

Guest Editorial: Laser Damage  

SciTech Connect

Laser damage of optical materials, first reported in 1964, continues to limit the output energy and power of pulsed and continuous-wave laser systems. In spite of some 48 years of research in this area, interest from the international laser community to laser damage issues remains at a very high level and does not show any sign of decreasing. Moreover, it grows with the development of novel laser systems, for example, ultrafast and short-wavelength lasers that involve new damage effects and specific mechanisms not studied before. This interest is evident from the high level of attendance and presentations at the annual SPIE Laser Damage Symposium (aka, Boulder Damage Symposium) that has been held in Boulder, Colorado, since 1969. This special section of Optical Engineering is the first one devoted to the entire field of laser damage rather than to a specific part. It is prepared in response to growing interest from the international laser-damage community. Some papers in this special section were presented at the Laser Damage Symposium; others were submitted in response to the general call for papers for this special section. The 18 papers compiled into this special section represent many sides of the broad field of laser-damage research. They consider theoretical studies of the fundamental mechanisms of laser damage including laser-driven electron dynamics in solids (O. Brenk and B. Rethfeld; A. Nikiforov, A. Epifanov, and S. Garnov; T. Apostolova et al.), modeling of propagation effects for ultrashort high-intensity laser pulses (J. Gulley), an overview of mechanisms of inclusion-induced damage (M. Koldunov and A. Manenkov), the formation of specific periodic ripples on a metal surface by femtosecond laser pulses (M. Ahsan and M. Lee), and the laser-plasma effects on damage in glass (Y. Li et al). Material characterization is represented by the papers devoted to accurate and reliable measurements of absorption with special emphasis on thin films (C. Mhlig and S. Bublitz; B. Cho, E. Danielewicz, and J. Rudisill; W. Palm et al; and J. Lu et al.). Statistical treatment of measurements of the laser-damage threshold (J. Arenberg) and the relationship to damage mechanisms (F. Wagner et al.) represent the large subfield of laser-damage measurements. Various aspects of multilayer coating and thin-film characterization are considered in papers by B. Cho, J. Rudisill, and E. Danielewicz (spectral shift in multilayer mirrors) and R. Weber et al. (novel approach to damage studies based on third-harmonic generation microscopy). Of special interest for readers is the paper by C. Stolz that summarizes the results of four thin-film damage competitions organized as a part of the Laser Damage Symposium. Another paper is devoted to thermal annealing of damage precursors (N. Shen et al.). Finally, the influence of nano-size contamination on initiation of laser damage by ultrashort pulses is considered in paper of V. Komolov et al.

Vitaly Gruzdev, Michelle D. Shinn

2012-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Web Based Course: SAF-230DE, Accident Investigation Overview...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video September 20, 2013 -...

222

ORISE: The Medical Basis for Radiation-Accident Preparedness...  

NLE Websites -- All DOE Office Websites (Extended Search)

The Medical Basis for Radiation-Accident Preparedness: Medical Management Proceedings of the Fifth International REACTS Symposium on the Medical Basis for Radiation-Accident...

223

Audit of the Department of Energy's Transportation Accident Resistant...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Transportation Accident Resistant Container Program, IG-0380 Audit of the Department of Energy's Transportation Accident Resistant Container Program, IG-0380 Audit of the...

224

Intermediate leak protection/automatic shutdown for B and W helical coil steam generator  

SciTech Connect

The report summarizes a follow-on study to the multi-tiered Intermediate Leak/Automatic Shutdown System report. It makes the automatic shutdown system specific to the Babcock and Wilcox (B and W) helical coil steam generator and to the Large Development LMFBR Plant. Threshold leak criteria specific to this steam generator design are developed, and performance predictions are presented for a multi-tier intermediate leak, automatic shutdown system applied to this unit. Preliminary performance predictions for application to the helical coil steam generator were given in the referenced report; for the most part, these predictions have been confirmed. The importance of including a cover gas hydrogen meter in this unit is demonstrated by calculation of a response time one-fifth that of an in-sodium meter at hot standby and refueling conditions.

1981-01-01T23:59:59.000Z

225

Saltwell Leak Detector Station Programmable Logic Controller (PLC) Software Configuration Management Plan (SCMP)  

Science Conference Proceedings (OSTI)

This document provides the procedures and guidelines necessary for computer software configuration management activities during the operation and maintenance phases of the Saltwell Leak Detector Stations as required by HNF-PRO-309, Rev. 1, Computer Software Quality Assurance, Section 2.4, Software Configuration Management. The software configuration management plan (SCMP) integrates technical and administrative controls to establish and maintain technical consistency among requirements, physical configuration, and documentation for the Saltwell Leak Detector Station Programmable Logic Controller (PLC) software during the Hanford application, operations and maintenance. This SCMP establishes the Saltwell Leak Detector Station PLC Software Baseline, status changes to that baseline, and ensures that software meets design and operational requirements and is tested in accordance with their design basis.

WHITE, K.A.

2000-11-28T23:59:59.000Z

226

SIXTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE  

SciTech Connect

A series of experiments to monitor the aging performance of Viton{reg_sign} GLT O-rings used in the Model 9975 package has been ongoing for seven years at the Savannah River National Laboratory. Seventy tests using mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 F. They were leak-tested initially and have been tested periodically to determine if they meet the criterion of leak-tightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 F. High temperature aging continues for 33 GLT O-ring fixtures at 200-300 F. Room temperature leak test failures have been experienced in all of the GLT O-ring fixtures aging at 350 F and higher temperatures, and in 7 fixtures aging at 300 F. No failures have yet been observed in GLT O-ring fixtures aging at 200 F for 41-60 months, which is still bounding to O-ring temperatures during storage in K-Area Complex (KAC). Based on expectations that the fixtures aging at 200 F will remain leak-tight for a significant period yet to come, 2 additional fixtures began aging within the past year at an intermediate temperature of 270 F, with hopes that they may leak before the 200 F fixtures. High temperature aging continues for 6 GLT-S O-ring fixtures at 200-300 F. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 F. No failures have yet been observed in GLT-S O-ring fixtures aging at 200-300 F for up to 26 months. For O-ring fixtures that have failed the room temperature leak test and been disassembled, the Orings displayed a compression set ranging from 51-96%. This is greater than seen to date for packages inspected during KAC field surveillance (24% average). For GLT O-rings, separate service life estimates have been made based on the O-ring fixture leak test data and based on compression stress relaxation (CSR) data. These two predictive models show reasonable agreement at higher temperatures (350-400 F). However, at 300 F, the room temperature leak test failures to date experienced longer aging times than predicted by the CSR-based model. This suggests that extrapolations of the CSR model predictions to temperatures below 300 F will provide a conservative prediction of service life relative to the leak rate criterion. Leak test failure data at lower temperatures are needed to verify this apparent trend. Insufficient failure data exist currently to perform a similar comparison for GLT-S O-rings. Aging and periodic leak testing will continue for the remaining fixtures.

Daugherty, W.

2011-08-31T23:59:59.000Z

227

MEMOIRS OF A LEAK: Infiltrating Research for a Quarter of a Century  

NLE Websites -- All DOE Office Websites (Extended Search)

MEMOIRS OF A LEAK: Infiltrating Research for a Quarter of a Century MEMOIRS OF A LEAK: Infiltrating Research for a Quarter of a Century Speaker(s): Max Sherman Date: November 16, 2000 - 12:00pm Location: Bldg. 90 Seminar Host/Point of Contact: David Faulkner Infiltration is the (usually uncontrolled) flow of air through leaks in the building envelope, driven by natural and mechanical pressures. Before the oil crises, there was not a lot of interest in infiltration. For houses and other envelope-dominated buildings, however, infiltration typically accounted for all of their ventilation needs and 1/3-1/2 of their space-conditioning load. Starting in the mid-70s there was a realization that this important problem was not well understood, but represented an important energy-saving opportunity. Research institutions around the world

228

Safety evaluation of MHTGR licensing basis accident scenarios  

SciTech Connect

The safety potential of the Modular High-Temperature Gas Reactor (MHTGR) was evaluated, based on the Preliminary Safety Information Document (PSID), as submitted by the US Department of Energy to the US Nuclear Regulatory Commission. The relevant reactor safety codes were extended for this purpose and applied to this new reactor concept, searching primarily for potential accident scenarios that might lead to fuel failures due to excessive core temperatures and/or to vessel damage, due to excessive vessel temperatures. The design basis accident scenario leading to the highest vessel temperatures is the depressurized core heatup scenario without any forced cooling and with decay heat rejection to the passive Reactor Cavity Cooling System (RCCS). This scenario was evaluated, including numerous parametric variations of input parameters, like material properties and decay heat. It was found that significant safety margins exist, but that high confidence levels in the core effective thermal conductivity, the reactor vessel and RCCS thermal emissivities and the decay heat function are required to maintain this safety margin. Severe accident extensions of this depressurized core heatup scenario included the cases of complete RCCS failure, cases of massive air ingress, core heatup without scram and cases of degraded RCCS performance due to absorbing gases in the reactor cavity. Except for no-scram scenarios extending beyond 100 hr, the fuel never reached the limiting temperature of 1600/degree/C, below which measurable fuel failures are not expected. In some of the scenarios, excessive vessel and concrete temperatures could lead to investment losses but are not expected to lead to any source term beyond that from the circulating inventory. 19 refs., 56 figs., 11 tabs.

Kroeger, P.G.

1989-04-01T23:59:59.000Z

229

SEVENTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE  

SciTech Connect

A series of experiments to monitor the aging performance of Viton GLT O-rings used in the Model 9975 package has been ongoing since 2004 at the Savannah River National Laboratory. Seventy tests using mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 F. They were leak-tested initially and have been tested periodically to determine if they meet the criterion of leak-tightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 F. High temperature aging continues for 23 GLT O-ring fixtures at 200 270 F. Room temperature leak test failures have been experienced in all of the GLT O-ring fixtures aging at 350 F and higher temperatures, and in 8 fixtures aging at 300 F. The remaining GLT O-ring fixtures aging at 300 F have been retired from testing following more than 5 years at temperature without failure. No failures have yet been observed in GLT O-ring fixtures aging at 200 F for 54-72 months, which is still bounding to O-ring temperatures during storage in K-Area Complex (KAC). Based on expectations that the fixtures aging at 200 F will remain leak-tight for a significant period yet to come, 2 additional fixtures began aging in 2011 at an intermediate temperature of 270 F, with hopes that they may reach a failure condition before the 200 F fixtures. High temperature aging continues for 6 GLT-S O-ring fixtures at 200 300 F. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 F. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 - 300 F for 30 - 36 months. For O-ring fixtures that have failed the room temperature leak test and been disassembled, the O-rings displayed a compression set ranging from 51 96%. This is greater than seen to date for any packages inspected during KAC field surveillance (24% average). For GLT O-rings, separate service life estimates have been made based on the O-ring fixture leak test data and based on compression stress relaxation (CSR) data. These two predictive models show reasonable agreement at higher temperatures (350 400 F). However, at 300 F, the room temperature leak test failures to date experienced longer aging times than predicted by the CSRbased model. This suggests that extrapolations of the CSR model predictions to temperatures below 300 F will provide a conservative prediction of service life relative to the leak rate criterion. Leak test failure data at lower temperatures are needed to verify this apparent trend. Insufficient failure data exist currently to perform a similar comparison for GLT-S O-rings. Aging and periodic leak testing will continue for the remaining PCV O-ring fixtures.

Daugherty, W.

2012-08-30T23:59:59.000Z

230

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Volume 6, Part 2: Appendices  

SciTech Connect

The objectives are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed. Volume 1 summarizes the results of the study. The scope of the level-1 study includes plant damage state analyses, and uncertainty analysis. The internal event analysis is documented in Volume 2. The internal fire and internal flood analysis are documented in Volumes 3 and 4, respectively. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associated, Inc. A phased approach was used in the level 2/3 PRA program, however both phases addressed the risk from only mid-loop operation. The first phase of the level 2/3 PRA was initiated in late 1991 and consisted of an Abridged Risk Study. This study was completed in May 1992 and was focused on accident progression and consequences, conditional on core damage. Phase 2 is a more detailed study in which an evaluation of risk during mid-loop operation was performed. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6. This report, Volume 6, Part 2, consists of five appendices containing supporting information for: the PDS (plant damage state) analysis; the accident progression analysis; the source term analysis; the consequence analysis; and the Melcor analysis. 73 figs., 21 tabs.

Jo, J.; Lin, C.C.; Neymotin, L.; Mubayi, V. [Brookhaven National Lab., Upton, NY (United States)

1995-05-01T23:59:59.000Z

231

Efficient model-based leak detection in boiler steam-water systems Xi Sun, Tongwen Chen *, Horacio J. Marquez  

E-Print Network (OSTI)

Efficient model-based leak detection in boiler steam-water systems Xi Sun, Tongwen Chen *, Horacio detection in boiler steam-water systems. The algorithm has been tested using real industrial data from Syncrude Canada, and has proven to be effective in detection of boiler tube or steam leaks; proper

Marquez, Horacio J.

232

Acetic Acid Sclerotherapy for Treatment of a Bile Leak from an Isolated Bile Duct After Laparoscopic Cholecystectomy  

Science Conference Proceedings (OSTI)

Bile leak after laparoscopic cholecystectomy is not uncommon, and it mainly occurs from the cystic duct stump and can be easily treated by endoscopic techniques. However, treatment for leakage from an isolated bile duct can be troublesome. We report a successful case of acetic acid sclerotherapy for bile leak from an isolated bile duct after laparoscopic cholecystectomy.

Choi, Gibok, E-mail: choigibok@yahoo.co.kr; Eun, Choong Ki, E-mail: ilovegod@chollian.net [Inje University, Department of Radiology, Haeundae Paik Hospital, College of Medicine (Korea, Republic of); Choi, HyunWook, E-mail: gdkid92@daum.net [Maryknoll Medical Center, Department of Radiology (Korea, Republic of)

2011-02-15T23:59:59.000Z

233

Study on New Methods of Improving the Accuracy of Leak Detection and Location of Natural Gas Pipeline  

Science Conference Proceedings (OSTI)

As negative pressure wave is applied to leak detection and location of natural gas pipeline, the key is how to realize accurate measurement of propagation velocity of pressure wave and time difference. However, there exists problem of lower accuracy ... Keywords: natural gas pipeline, leak detection and location, negative pressure wave, wavelet transform, singularity detection

Shuqing Zhang; Tianye Gao; Hong Xu; Guangpu Hao; Zhongdong Wang

2009-04-01T23:59:59.000Z

234

Decontamination Dressdown at a Transportation Accident Involving  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Decontamination Dressdown at a Transportation Accident Involving Decontamination Dressdown at a Transportation Accident Involving Radioactive Material Decontamination Dressdown at a Transportation Accident Involving Radioactive Material The purpose of this User's Guide is to provide instructors with an overview of the key points covered in the video. The Student Handout portion of this Guide is designed to assist the instructor in reviewing those points with students. The Student Handout should be distributed to students after the video is shown and the instructor should use the Guide to facilitate a discussion on how the decontamination dressdown process is implemented. During this discussion, the instructor can present various scenarios, each of which would discuss decontamination at the accident scene. The purpose of this discussion would be to cover how responders

235

Chernobyl accident: A comprehensive risk assessment  

SciTech Connect

The authors, all of whom are Ukrainian and Russian scientists involved with Chernobyl nuclear power plant since the April 1986 accident, present a comprehensive review of the accident. In addition, they present a risk assessment of the remains of the destroyed reactor and its surrounding shelter, Chernobyl radioactive waste storage and disposal sites, and environmental contamination in the region. The authors explore such questions as the risks posed by a collapse of the shelter, radionuclide migration from storage and disposal facilities in the exclusion zone, and transfer from soil to vegetation and its potential regional impact. The answers to these questions provide a scientific basis for the development of countermeasures against the Chernobyl accident in particular and the mitigation of environmental radioactive contamination in general. They also provide an important basis for understanding the human health and ecological risks posed by the accident.

Vargo, G.J.; Poyarkov, V.; Baryakhtar, V.; Kukhar, V.; Los, I.

1999-11-01T23:59:59.000Z

236

A systems approach to food accident analysis  

E-Print Network (OSTI)

Food borne illnesses lead to 3000 deaths per year in the United States. Some industries, such as aviation, have made great strides increasing safety through careful accident analysis leading to changes in industry practices. ...

Helferich, John D

2011-01-01T23:59:59.000Z

237

Oil spill nears the beaches of Florida, and the leak may not be plugged before Christmas  

E-Print Network (OSTI)

Oil spill nears the beaches of Florida, and the leak may not be plugged before Christmas By David Gardner Last updated at 11:32 AM on 3rd June 2010 BP's giant oil slick was bearing down on Florida holidaymakers a year visit Florida and state leaders fear the oil will devastate a tourist industry

Belogay, Eugene A.

238

In Shop Acceptance Test Report for the SY Farm Annulus Leak Detectors  

SciTech Connect

The following test report was written for the SY tank farm annulus leak detectors. The test plan used was HNF-4546, Revision 1. The purpose of the test plan was to test the ENRAF series 854 ATG with SPU II card prior to installation. The test plan set various parameters and verifies the gauge and alarms functionality.

SMITH, S.G.

1999-12-07T23:59:59.000Z

239

Application of the Leak-Before-Break Approach to BWR Piping  

Science Conference Proceedings (OSTI)

By applying the leak-before-break approach in BWR safety evaluations, utilities can justify eliminating many pipe whip restraints and jet impingement shields. The report details a sample analysis of recirculation piping in a General Electric BWR and offers a procedure for ranking plant piping systems for analysis.

1987-01-07T23:59:59.000Z

240

Leaking electricity: Standby and off-mode power consumption in consumer electronics and household appliances  

Science Conference Proceedings (OSTI)

This report assesses ``leaking`` electricity from consumer electronics and small household appliances when they are in standby mode or turned off, and examines the impacts of these losses. The report identifies trends in relevant product industries and gives technical and policy options for reducing standby and off-mode power loss.

Thorne, J.; Suozzo, M.

1998-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Tank 241-AY-102 Leak Assessment Supporting Documentation: Miscellaneous Reports, Letters, Memoranda, And Data  

SciTech Connect

This report contains reference materials cited in RPP-ASMT -53793, Tank 241-AY-102 Leak Assessment Report, that were obtained from the National Archives Federal Records Repository in Seattle, Washington, or from other sources including the Hanford Site's Integrated Data Management System database (IDMS).

Engeman, J. K.; Girardot, C. L.; Harlow, D. G.; Rosenkrance, C. L.

2012-12-20T23:59:59.000Z

242

A SUMMARY OF INDUSTRIAL ACCIDENTS IN USAEC FACILITIES  

SciTech Connect

The summary includes descriptions of serious accidents for l959 and 1960, AEC industrial injury frequency rates, criticality accidents, radiation exposures, accidents involving radioactive materials in AEC activities during 1959 and 1960, and accidents involving fatalities in AEC activities during l959 and 1960. (B.O.G.)

1961-12-01T23:59:59.000Z

243

Three Mile Island accident and post-accident recovery: what did we learn  

SciTech Connect

A description of the accident at Three Mile Island-2 reactor is presented. Activities related to the cleanup and decontamination of the reactor are described.

Collins, E.D.

1982-01-01T23:59:59.000Z

244

FIFTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE  

Science Conference Proceedings (OSTI)

A series of experiments to monitor the aging performance of Viton{sup reg.} GLT O-rings used in the Model 9975 package has been ongoing for six years at the Savannah River National Laboratory. Sixty-seven mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 F. They were leak-tested initially and have been tested at nominal six month intervals to determine if they meet the criterion of leaktightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 F. High temperature aging continues for 36 GLT O-ring fixtures at 200--350 F. Room temperature leak test failures have been experienced in 5 of the GLT O-ring fixtures aging at 300 and 350 F, and in all 3 of the GLT O-ring fixtures aging at higher temperatures. No failures have yet been observed in GLT O-ring fixtures aging at 200 F for 30--48 months, which is still bounding to O-ring temperatures during storage in KAMS. High temperature aging continues for 6 GLT-S O-ring fixtures at 200--300 F. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 F. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 or 300 F for 19 months. For O-ring fixtures that have failed the room temperature leak test and been disassembled, the O-rings displayed a compression set ranging from 51--95%. This is significantly greater than seen to date for packages inspected during KAMS field surveillance (23% average). For GLT O-rings, service life based on the room temperature leak rate criterion is comparable to that predicted by compression stress relaxation (CSR) data at higher temperatures (350--400 F). While there are no comparable failure data yet at aging temperatures below 300 F, extrapolations of the data for GLT O-rings suggests that CSR model predictions provide a conservative prediction of service life relative to the leak rate criterion. Failure data at lower temperatures is needed to verify this apparent trend. Insufficient failure data exist currently to perform a similar comparison for GLT-S O-rings. Aging and periodic leak testing will continue for the remaining fixtures.

Daugherty, W.; Hoffman, E.

2010-11-01T23:59:59.000Z

245

FIFTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE  

SciTech Connect

A series of experiments to monitor the aging performance of Viton{reg_sign} GLT O-rings used in the Model 9975 package has been ongoing for six years at the Savannah River National Laboratory. Sixty-seven mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 F. They were leak-tested initially and have been tested at nominal six month intervals to determine if they meet the criterion of leaktightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 F. High temperature aging continues for 36 GLT O-ring fixtures at 200-350 F. Room temperature leak test failures have been experienced in 6 of the GLT O-ring fixtures aging at 300 and 350 F, and in all 3 of the GLT O-ring fixtures aging at higher temperatures. No failures have yet been observed in GLT O-ring fixtures aging at 200 F for 30-48 months, which is still bounding to O-ring temperatures during storage in KAMS. High temperature aging continues for 6 GLT-S O-ring fixtures at 200-300 F. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 F. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 or 300 F for 19 months. For O-ring fixtures that have failed the room temperature leak test and been disassembled, the Orings displayed a compression set ranging from 51-95%. This is significantly greater than seen to date for packages inspected during KAMS field surveillance (23% average). For GLT O-rings, service life based on the room temperature leak rate criterion is comparable to that predicted by compression stress relaxation (CSR) data at higher temperatures (350-400 F). While there are no comparable failure data yet at aging temperatures below 300 F, extrapolations of the data for GLT O-rings suggests the CSR model predictions provide a conservative prediction of service life relative to the leak rate criterion. Failure data at lower temperatures are needed to verify this apparent trend. Insufficient failure data exist currently to perform a similar comparison for GLT-S O-rings. Aging and periodic leak testing will continue for the remaining fixtures.

Daugherty, W.; Hoffman, E.

2011-04-11T23:59:59.000Z

246

Cobalt-60 simulation of LOCA (loss of coolant accident) radiation effects  

Science Conference Proceedings (OSTI)

The consequences of simulating nuclear reactor loss of coolant accident (LOCA) radiation effects with Cobalt-60 gamma ray irradiators have been investigated. Based on radiation induced damage in polymer base materials, it was demonstrated that electron/photon induced radiation damage could be related on the basis of average absorbed radiation dose. This result was used to estimate the relative effectiveness of the mixed beta/gamma LOCA and Cobalt-60 radiation environments to damage both bare and jacketed polymer base electrical insulation materials. From the results obtained, it is concluded that present simulation techniques are a conservative method for simulating LOCA radiation effects and that the practices have probably substantially overstressed both bare and jacketed materials during qualification testing. 9 refs., 8 figs., 5 tabs.

Buckalew, W.H.

1989-07-01T23:59:59.000Z

247

Discovery of the First Leaking Double-Shell Tank - Hanford Tank 241-AY-102  

SciTech Connect

A routine video inspection of the annulus region of double-shell tank 241-A Y-102 in August of 2012 indicated the presence material in the annulus space between the primary and secondary liners. A comparison was made to previous inspections perfom1ed in 2006 and 2007. which indicated that a change had occurred. The material was observed at two locations on the floor of the annulus and one location at the top of the annulus region where the primary and secondary top knuckles meet (RPP-ASMT-53793). Subsequent inspections were performed. leading to additional material observed on the floor of the annulus space in a region that had not previously been inspected (WRPS-PER-2012-1363). The annulus Continuous Air Monitor (CAM) was still operational and was not indicating elevated radiation levels in the annulus region. When the camera from the inspections was recovered. it also did not indicate increased radiation above minimum contamination levels (WRPS-PER-2012-1363). A formal leak assessment team was established August 10, 2012 to review tank 241-AY-102 construction and operating histories and to determine whether the material observed in the annulus had resulted from a leak in the primary tank. The team consisted of individuals from Engineering. Base Operations.and Environmental Protection. As this was a first-of-its-kind task. a method for obtaining a sample of the material in the annulus was needed. The consistency of the material was unknown.and the location of a majority of the material was not conducive to using the sampling devices that were currently available at Hanford. A subcontractor was tasked with the development fabrication.and testing of a sampling device that would be able to obtain multiple samples from the material on the annulus floor. as well as the material originating from a refractory air-slot near the floor of the annulus space. This sampler would need to be able to collect and dispense the material it collected into a sample jar retrieval device for transportation of the material to the 222-S laboratory on the Hanford site for analysis. The subcontractor agency fabricated a remote underground sampler by modifying off-the-shelf robotics and parts. Limited testing of the sampler was conducted using a mock-up of the tank annulus and one simulated material type -a salt block. The mock-up testing indicated that the sampler would be able to maneuver within the confined space and that the device worked with full functionality. A total of six weeks had passed from initiation to implementation of the new sampler in the 241-AY-102 tank annulus. Initial sample material was obtained from the annulus floor using the Off-Riser Sampler System that has been used at Hanford tor years to obtain material from the primary tanks. This could be used at the location near Riser 83 since the material was collected directly from the annulus floor and not from a location on the wall or behind a pipe, as was needed from the two locations near Riser 90. After obtaining a small sample of the material on the annulus floor.this sampler sustained terminal damage due to conduit pipes it had to transverse in order to collect and recover material from this location. Several issues were also encountered during deployment of the new sampler into the annulus near Riser 90. These included: Difficulty fitting the sampler down the 12-inch riser into the annulus due to a small tolerance in the size ofthe sampler Failure of sampler components and functions during deployment including the camera. pneumatics.and bearing seals Delays in the field due to supporting equipment issues including cables. cameras. and scaffolding Low recovery of sample material obtained for analysis The complications that occurred during deployment and use of the new sampler during the sampling event ultimately resulted in lower recovery of material from these locations in the annulus than was obtained using the Off-Riser Sampler System and limited the analyses that could be performed for determining the origin of the material. Following completion of th

Harrington, Stephanie J.; Sams, Terry L.

2013-08-15T23:59:59.000Z

248

Model based detection of hydrogen leaks in a fuel cell stack Ari Ingimundarson and Anna G. Stefanopoulou and Denise McKay  

E-Print Network (OSTI)

will depend on the composition of the gas where the leak takes place. Two approaches are presented here but takes into account the natural leak of the stack and humidity. Hydrogen leak detection without using. Hydrogen has the lowest molecular weight and viscosity of any gas. Its properties make it have a faster

Stefanopoulou, Anna

249

Primary Radiation Damage Formation  

SciTech Connect

The physical processes that give rise to changes in the microstructure, and the physical and mechanical properties of materials exposed to energetic particles are initiated by essentially elastic collisions between atoms in what has been called an atomic displacement cascade. The formation and evolution of this primary radiation damage mechanism are described to provide an overview of how stable defects are formed by displacement cascades, as well as the nature and morphology of the defects themselves. The impact of the primary variables cascade energy and irradiation temperature are discussed, along with a range of secondary factors that can influence damage formation.

Stoller, Roger E [ORNL

2012-01-01T23:59:59.000Z

250

Instrument Performance Under Severe Accident Conditions: Ways to Acquire Information From Instrumentation Affected by an Accident  

Science Conference Proceedings (OSTI)

Under accident conditions, information is needed for diagnosing plant status and confirming plant responses to mitigative actions. This makes it important to understand how instruments behave in severe accident environments and to find ways to obtain information from the instruments under conditions that can be more severe than their design bases.

1993-12-01T23:59:59.000Z

251

Possible Methods to Estimate Core Location in a Beyond-Design-Basis Accident at a GE BWR with a Mark I Containment Stucture  

SciTech Connect

It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting in a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.

Walston, S; Rowland, M; Campbell, K

2011-07-27T23:59:59.000Z

252

SAF-230DE - Web Based Course: Accident Investigation Overview | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SAF-230DE - Web Based Course: Accident Investigation Overview SAF-230DE - Web Based Course: Accident Investigation Overview SAF-230DE - Web Based Course: Accident Investigation Overview September 18, 2013 - 10:52am Addthis SAF-230DE - Web Based Course: Accident Investigation Overview The Office of Health Safety and Security (HSS) National Training Center (NTC) in collaboration with the HSS Accident Investigation Program (HS-24) has developed and released a course that provides an overview of the fundamentals of accident investigation. This course is intended to meet the every five year refresher training requirement for DOE Federal Accident Investigators under DOE Order 225.1B "Accident Investigations", and serves as an orientation to other DOE Federal Accident Investigation Board Members who need a basic knowledge of

253

Accident source terms for boiling water reactors with high burnup cores.  

Science Conference Proceedings (OSTI)

The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

2007-11-01T23:59:59.000Z

254

Nuclear Maintenance Applications Center: On-Line Leak Sealing: A Guide for Nuclear Power Plant Maintenance Personnel (Update of NP-6 523-D)  

Science Conference Proceedings (OSTI)

On-line leak sealing consists of repairing a leak or potential leak while the plant is operating. A pre-engineered fixture or part of the component itself is used to form a cavity that will contain the leak source. The fixture design includes a means of injecting the sealant using a shutoff adapter. The injection equipment is attached to the adaptor, and sealant is injected either directly into the cavity or into a peripheral seal to seal the leak. Although the methodology has been in existence for 45 ye...

2011-06-29T23:59:59.000Z

255

Is the situation and immediate threat to life and health? Spill/Leak/Release Medical Emergency Fire or Flammable Gas Spill/Leak/Release Medical Emergency Fire or Flammable Gas Chemical Odor? Possible Fire / Natural Gas  

E-Print Network (OSTI)

? Possible Fire / Natural Gas (including chemicals and bio agents") (not including chemicals or bio agents Fire or Flammable Gas Spill/Leak/Release Medical Emergency Fire or Flammable Gas Chemical Odor

256

EMERGENCY RESPONSE TO A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Emer Emer Emer Emer Emer Emergency Response to a T gency Response to a T gency Response to a T gency Response to a T gency Response to a Transportation ransportation ransportation ransportation ransportation Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material Accident Involving Radioactive Material DISCLAIMER DISCLAIMER DISCLAIMER DISCLAIMER DISCLAIMER Viewing this video and completing the enclosed printed study material do not by themselves provide sufficient skills to safely engage in or perform duties related to emergency response to a transportation accident involving radioactive material. Meeting that goal is beyond the scope of this video and requires either additional

257

A Review of Criticality Accidents 2000 Revision  

SciTech Connect

Criticality accidents and the characteristics of prompt power excursions are discussed. Sixty accidental power excursions are reviewed. Sufficient detail is provided to enable the reader to understand the physical situation, the chemistry and material flow, and when available the administrative setting leading up to the time of the accident. Information on the power history, energy release, consequences, and causes are also included when available. For those accidents that occurred in process plants, two new sections have been included in this revision. The first is an analysis and summary of the physical and neutronic features of the chain reacting systems. The second is a compilation of observations and lessons learned. Excursions associated with large power reactors are not included in this report.

Thomas P. McLaughlin; Shean P. Monahan; Norman L. Pruvost; Vladimir V. Frolov; Boris G. Ryazanov; Victor I. Sviridov

2000-05-01T23:59:59.000Z

258

Severe accident testing of a personnel airlock  

Science Conference Proceedings (OSTI)

Sandia National Laboratories (Sandia) is investigating the leakage potential of mechanical penetrations as part of a research program on containment integrity under severe accident loads for the US Nuclear Regulatory Commission (NRC). Barnes et al. (1984) and Shackelford et al. (1985) identified leakage from personnel airlocks as an important failure mode of containments subject to severe accident loads. However, these studies were based on relatively simple analysis methods. The complex structural interaction between the door, gasket, and bulkhead in personnel airlocks makes analytical evaluation of leakage difficult. In order to provide data to validate methods for evaluating the leakage potential, a full-size personnel airlock was subject to simulated severe accident loads consisting of pressure and temperature up to 300 psig and 800/degree/F. The test was conducted at Chicago Bridge and Iron under contract to Sandia. Julien and Peters (1989) provide a detailed report on the test program. 6 refs., 5 figs.

Clauss, D.B.; Parks, M.B.; Julien, J.T.; Peters, S.W.

1989-01-01T23:59:59.000Z

259

Assessment of CRBR core disruptive accident energetics  

Science Conference Proceedings (OSTI)

The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly.

Theofanous, T.G.; Bell, C.R.

1984-03-01T23:59:59.000Z

260

Analysis of Severe Accident Management Strategy for a BWR-4 Nuclear Power Plant  

SciTech Connect

The Chinshan nuclear power plant (NPP) is a Mark-I boiling water reactor (BWR) NPP located in northern Taiwan. The Chinshan NPP severe accident management guidelines (SAMGs) were developed based on the BWR Owners Group Emergency Procedure Guidelines/Severe Accident Guidelines and were developed at the end of 2003. The MAAP4 code has been used as a tool to validate the SAMG strategies. The development process and characteristics of the Chinshan SAMGs are described. The T{sub 5}U{sub t}X{sub C} sequence, the highest core damage frequency in the probabilistic risk assessment insight of the Chinshan NPP, is cited as a reference case for SAMG validation. Not all safety injection systems are operated in the T{sub 5}U{sub t}X{sub C} sequence. The severe accident progression is simulated, and the entry condition of the SAMGs is described. Then, the T{sub 5}U{sub t}X{sub C} sequence is simulated with reactor pressure vessel (RPV) depressurization. Mitigation actions based on the Chinshan NPP SAMGs are then applied to demonstrate the effectiveness of the SAMGs. Sensitivity studies on RPV depressurization with the reactor water level and minimum RPV injection flow rate are also investigated in this study. Based on MAAP4 calculation and the default values of the parameters calculating the severe accident phenomena, the result shows that RPV depressurization before the reactor water level reaches one-fourth of the core water level can prevent the core from damage in the T{sub 5}U{sub t}X{sub C} sequence. The flow rate of two control rod drive pumps is enough to maintain the reactor water level above the top of active fuel and cool down the core in the T{sub 5}U{sub t}X{sub C} sequence without operator action.

Wang, T.-C.; Wang, S.-J.; Teng, J.-T

2005-12-15T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

ENVIRONMENTAL MONITORING OF LEAKS USING TIME LAPSED LONG ELECTRODE ELECTRICAL RESISTIVITY  

Science Conference Proceedings (OSTI)

Highly industrialized areas pose challenges for surface electrical resistivity characterization due to metallic infrastructure. The infrastructure is typically more conductive than the desired targets and will mask the deeper subsurface information. These challenges may be minimized if steel-cased wells are used as long electrodes in the area near the target. We demonstrate a method of using long electrodes to electrically monitor a simulated leak from an underground storage tank with both synthetic examples and a field demonstration. The synthetic examples place a simple target of varying electrical properties beneath a very low resistivity layer. The layer is meant to replicate the effects of infrastructure. Both surface and long electrodes are tested on the synthetic domain. The leak demonstration for the field experiment is simulated by injecting a high conductivity fluid in a perforated well within the S tank farm at Hanford, and the resistivity measurements are made before and after the leak test. All data are processed in four dimensions, where a regularization procedure is applied in both the time and space domains. The synthetic test case shows that the long electrode ERM could detect relative changes in resistivity that are commensurate with the differing target properties. The surface electrodes, on the other hand, had a more difficult time matching the original target's footprint. The field results shows a lowered resistivity feature develop south of the injection site after cessation of the injections. The time lapsed regularization parameter has a strong influence on the differences in inverted resistivity between the pre and post injection datasets, but the interpretation of the target is consistent across all values of the parameter. The long electrode ERM method may provide a tool for near real-time monitoring of leaking underground storage tanks.

MYERS DA; RUCKER DF; FINK JB; LOKE MH

2009-12-16T23:59:59.000Z

262

LESSONS LEARNED FROM A RECENT LASER ACCIDENT  

SciTech Connect

A graduate student received a laser eye injury from a femtosecond Ti:sapphire laser beam while adjusting a polarizing beam splitter optic. The direct causes for the accident included failure to follow safe alignment practices and failure to wear the required laser eyewear protection. Underlying root causes included inadequate on-the-job training and supervision, inadequate adherence to requirements, and inadequate appreciation for dimly visible beams outside the range of 400-700nm. This paper describes how the accident occurred, discusses causes and lessons learned, and describes corrective actions being taken.

Woods, Michael; /SLAC

2011-01-26T23:59:59.000Z

263

Discovery of the First Leaking Double-Shell Tank - Hanford Tank 241-AY-102-14222  

SciTech Connect

A routine video inspection of the annulus space between the primary tank and secondary liner of double-shell tank 241-AY-102 was performed in August 2012. During the inspection, unexpected material was discovered. A subsequent video inspection revealed additional unexpected material on the opposite side of the tank, none of which had been observed during inspections performed in December 2006 and January 2007. A formal leak assessment team was established to review the tank's construction and operating histories, and preparations for sampling and analysis began to determine the material's origin. A new sampling device was required to collect material from locations that were inaccessible to the available sampler. Following its design and fabrication, a mock-up test was performed for the new sampling tool to ensure its functionality and capability of performing the required tasks. Within three months of the discovery of the unexpected material, sampling tools were deployed, material was collected, and analyses were performed. Results indicated that some of the unknown material was indicative of soil, whereas the remainder was consistent with tank waste. This, along with the analyses performed by the leak assessment team on the tank's construction history, lead to the conclusion that the primary tank was leaking into the annulus. Several issues were encountered during the deployment of the samplers into the annulus. As this was the first time samples had been required from the annulus of a double-shell tank, a formal lessons learned was created concerning designing equipment for unique purposes under time constraints.

Harrington, Stephanie J.; Sams, Terry L.

2013-11-06T23:59:59.000Z

264

CHARACTERIZATION OF LEAK PATHWAYS IN THE BELOW GRADE DUCTS OF THE BROOKHAVEN GRAPHITE RESEARCH REACTOR USING PERFLUOROCARBON TRACERS.  

SciTech Connect

The focus of this program was the characterization of the soils beneath the main air ducts connecting the exhaust plenums with the Fan House. The air plenums experienced water intrusion during BGRR operations and after shutdown. The water intrusions were attributed to rainwater leaks into degraded parts of the system and to internal cooling water system leaks. As part of the overall characterization efforts, a state-of-the-art gaseous perfluorocarbon tracer technology was utilized to characterize leak pathways from the ducts. This in turn suggests what soil regions under or adjacent to the ductwork should be emphasized in the characterization process. Knowledge of where gaseous tracers leak from the ducts yields a conservative picture of where water transport, out of or into, the ducts might have occurred.

HEISER,J.; SULLIVAN,T.; KALB,P.; MILIAN,L.; WILKE,R.; NEWSON,C.; LILIMPAKIS,M.

2001-04-01T23:59:59.000Z

265

Analysis of core damage frequency from internal events: Peach Bottom, Unit 2  

SciTech Connect

This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis.

Kolaczkowski, A.M.; Lambright, J.A.; Ferrell, W.L.; Cathey, N.G.; Najafi, B.; Harper, F.T.

1986-10-01T23:59:59.000Z

266

EIGHTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE  

SciTech Connect

A series of experiments to monitor the aging performance of Viton GLT O-rings used in the Model 9975 package has been ongoing since 2004 at the Savannah River National Laboratory. Seventy tests using mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 F. They were leak-tested initially and have been tested periodically to determine if they meet the criterion of leak-tightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 F. High temperature aging continues for 23 GLT O-ring fixtures at 200 270 F. Room temperature leak test failures have been experienced in all of the GLT O-ring fixtures aging at 350 F and higher temperatures, and in 8 fixtures aging at 300 F. The remaining GLT O-ring fixtures aging at 300 F have been retired from testing following more than 5 years at temperature without failure. No failures have yet been observed in GLT O-ring fixtures aging at 200 F for 61 - 85 months, which is still bounding to O-ring temperatures during storage in KArea Complex (KAC). Based on expectations that the fixtures aging at 200 F will remain leaktight for a significant period yet to come, 2 additional fixtures began aging in 2011 at an intermediate temperature of 270 F, with hopes that they may reach a failure condition before the 200 F fixtures. High temperature aging continues for 6 GLT-S O-ring fixtures at 200 300 F. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 F. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 - 300 F for 41 - 45 months. Aging and periodic leak testing will continue for the remaining PCV fixtures.

Daugherty, W. L.

2013-09-03T23:59:59.000Z

267

Development of the severe accident management guidelines (SAMG) for Ulchin Nuclear Power Plant Unit 3, 4, 5 and 6  

SciTech Connect

This paper describes the development process of the severe accident management guidelines (SAMG) for Units 3, 4, 5 and 6 of Ulchin Nuclear Power Plant. The units are Korean Standard Nuclear Power (KSNP) plant, 1000 MWe class pressurized water reactor (PWR) with two loops of primary coolant system. The severe accident management guidelines for the units have been completed in 2002. The generic severe accident management guidance for Korean Standard Nuclear Power Plant has been used as the basis when developing Ulchin severe accident management guideline. Result of probabilistic safety assessment (PSA) for each unit was reviewed to integrate its insight into the SAMG. It indicates that each unit has a balanced design to any specific initiating events for core damage. Seven severe accident management strategies are applied in Ulchin SAMG. Seven strategies are (1) Inject into the steam generator (2) De-pressurize the RCS (3) Inject into the RCS (4) Inject into the containment (5) Control the fission product release into environment (6) Control the containment pressure and temperature and (7) Control hydrogen concentration in the containment. The range and capability of essential instrument for performing the strategies are assessed. Computational aids are developed to complement the unavailable instrument during the accident and to assist the operator's decision choosing strategies. To examine the ability of the SAMG to fulfill its intended function, small loss of coolant accident (SLOCA) with the failure of safety injection was selected as a reference scenario. The scenario was analyzed using MAAP code. The evaluation of the SAMG using this sequence has been successfully completed. (authors)

Kim, Hyeong T.; Yoo, Hojong; Lim, Hyuk Soon; Park, Jong W.; Lim, Woosang; Oh, Seung Jong [Korea Hydro and Nuclear Power Co., Ltd., 103-16 Munji-Dong, Yusung-Gu, Daejeon, 305-380 (Korea, Republic of); Chung, Chang Hyun [Seoul National University (Korea, Republic of); Lee, Byung Chul [Future and Challenges, Inc (Korea, Republic of)

2004-07-01T23:59:59.000Z

268

Accident Investigation of the Fall Injury at the Savannah River...  

NLE Websites -- All DOE Office Websites (Extended Search)

U.S. Department of Energy Office of Environmental Management Accident Investigation Report Fall Injury Accident at the Savannah River Site on July 1, 2011 August 8, 2011 Disclaimer...

269

Uncertainty Assessments in Severe Nuclear Accident Scenarios  

Science Conference Proceedings (OSTI)

Managing uncertainties in industrial systems is a daily challenge to ensure improved design, robust operation, accountable performance and responsive risk control. This paper aims to illustrate the different depth analyses that the uncertainty software ... Keywords: Monte Carlo simulation, nuclear power plant, sensitivity analysis, severe accident, uncertainty

Bertrand Iooss; Fabrice Gaudier; Michel Marques; Bertrand Spindler; Bruno Tourniaire

2009-09-01T23:59:59.000Z

270

Blasting practices and explosives accidents in Utah coal mines  

SciTech Connect

Practices in use in Utah are commended and accidents incident to blasting are reviewed with suggestions as to future avoidance.

Parker, D.J.

1935-01-01T23:59:59.000Z

271

RECENT LASER ACCIDENTS AT DEPARTMENT OF ENERGY LABORATORIES  

SciTech Connect

Recent laser accidents and incidents at research laboratories across the Department of Energy complex are reviewed in this paper. Factors that contributed to the accidents are examined. Conclusions drawn from the accident reports are summarized and compared. Control measures that could have been implemented to prevent the accidents will be summarized and compared. Recommendations for improving laser safety programs are outlined and progress toward achieving them are summarized.

ODOM, CONNON R. [Los Alamos National Laboratory

2007-02-02T23:59:59.000Z

272

Accident Investigation and Materials Failure Analysis at the ...  

Science Conference Proceedings (OSTI)

Both are independent federal agencies charged with investigating transportation accidents in all modes, including aviation, railroad, highway, marine, pipeline,...

273

DOE O 225.1B, Accident Investigations  

Directives, Delegations, and Requirements

This Order prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, ...

2011-03-04T23:59:59.000Z

274

Market Damages and the Economic Waste Fallacy  

E-Print Network (OSTI)

Robert E. Scott, The Case for Market Damages; Revisiting thethey pay less than full market damages but when buyerssellers recover full market damages. As a consequence,

Scott, Robert E.; Schwartz, Alan

2008-01-01T23:59:59.000Z

275

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Main report. Volume 6. Part 1  

SciTech Connect

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. A phased approach was used in the level-1 program. In phase 1 which was completed in Fall 1991, a coarse screening analysis including internal fire and flood was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6.

Jo, J.; Lin, C.C.; Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)] [and others

1995-05-01T23:59:59.000Z

276

Environment/Health/Safety (EHS): Monthly Accident Statistics  

NLE Websites -- All DOE Office Websites (Extended Search)

Monthly Accident Statistics Monthly Accident Statistics Latest Accident Statistics Accident Statistics (through December 2013) Archived Accident Statistics 2013 Through November Through October Through September Through August Through July Through June Through May Through April Through March Through February Through January 2012 Through December Through November Through October Through September Through August Through July Through June Through May Through February Through January 2011 Through December Through November Through October Through September Through August Through July Through June Through May Through April Through March Through February Through January 2010 Through December Through November Through October Through September Through August Through July Through June Through May Through April Through March Through February

277

Trends status: Post-accident fission product chemistry  

DOE Green Energy (OSTI)

It is important to understand and model the chemical and physical behavior of vapor iodine species in containment environments for the following reasons: This behavior can contribute significantly to severe accident source terms; the development of accident mitigation or management strategies (e.g., an effective filter system); for long-term clean-up and recovery following an accident; regulatory requirements (e.g., spray or pool additives); and design basis accidents (i.e., steam generator tube rupture). This document discusses the Oak Ridge National Laboratory ''Post-Accident'' Chemistry Program.

Kress, T.S.; Beahm, E.C.; Shockley, W.C.; Weber, C.F.

1988-01-01T23:59:59.000Z

278

REAC/TS Radiation Accident Registry: An Overview  

Science Conference Proceedings (OSTI)

Over the past four years, REAC/TS has presented a number of case reports from its Radiation Accident Registry. Victims of radiological or nuclear incidents must meet certain dose criteria for an incident to be categorized as an accident and be included in the registry. Although the greatest numbers of accidents in the United States that have been entered into the registry involve radiation devices, the greater percentage of serious accidents have involved sealed sources of one kind or another. But if one looks at the kinds of accident scenarios that have resulted in extreme consequence, i.e., death, the greater share of deaths has occurred in medical settings.

Doran M. Christensen, DO, REAC /TS Associate Director and Staff Physician Becky Murdock, REAC/TS Registry and Health Physics Technician

2012-12-12T23:59:59.000Z

279

Leaking method approach to surface transport in the Mediterranean Sea from a numerical ocean model  

E-Print Network (OSTI)

We use Lagrangian diagnostics (the leaking and the exchange methods) to characterize surface transport out of and between selected regions in the Western Mediterranean. Velocity fields are obtained from a numerical model. Residence times of water of Atlantic origin in the Algerian basin, with a strong seasonal dependence, are calculated. Exchange rates between these waters and the ones occupying the northern basin are also evaluated. At surface, northward transport is dominant, and involves filamental features and eddy structures that can be identified with the Algerian eddies. The impact on these results of the presence of small scale turbulent motions is evaluated by adding Lagrangian diffusion.

Judit Schneider; Vicente Fernandez; Emilio Hernandez-Garcia

2004-10-01T23:59:59.000Z

280

Depressurization as an accident management strategy to minimize the consequences of direct containment heating  

Science Conference Proceedings (OSTI)

Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool  

SciTech Connect

MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ``MELCOR Verification, Benchmarking, and Applications,`` whose aim is to provide independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool. The scope of this program is to perform quality control verification on all released versions of MELCOR, to benchmark MELCOR against more mechanistic codes and experimental data from severe fuel damage tests, and to evaluate the ability of MELCOR to simulate long-term severe accident transients in commercial LWRs, by applying the code to model both BWRs and PWRs. Under this program, BNL provided input to the NRC-sponsored MELCOR Peer Review, and is currently contributing to the MELCOR Cooperative Assessment Program (MCAP). This paper presents a summary of MELCOR assessment efforts at BNL and their contribution to NRC goals with respect to MELCOR.

Madni, I.K. [Brookhaven National Lab., Upton, NY (United States); Eltawila, F. [Nuclear Regulatory Commission, Washington, DC (United States)

1994-01-01T23:59:59.000Z

282

Accident, Maryland: Energy Resources | Open Energy Information  

Open Energy Info (EERE)

Accident, Maryland: Energy Resources Accident, Maryland: Energy Resources Jump to: navigation, search Equivalent URI DBpedia Coordinates 39.628696°, -79.319759° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":39.628696,"lon":-79.319759,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

283

Characterization of a nuclear accident dosimeter  

E-Print Network (OSTI)

The 23rd nuclear accident dosimetry intercomparison was held during the week of June 12-16, 1995 at Los Alamos National Laboratory. This report presents the results of this event, referred to as NAD 23, as related to the performance of Sandia National Laboratories' (SNL) personal nuclear accident dosimeter (PNAD). Two separate critical assemblies, SHEBA and Godiva, were used to generate seven separate neutron spectra for use in dose comparisons. SNL's PNAD measured absorbed doses that were within +16 to +26 percent of the reference doses. In addition, a preliminary investigation was undertaken to determine the feasibility of using the data obtained from an irradiated PNAD to correct for body orientation. This portion of the experiment was performed with a TRIGA reactor at the Nuclear Science Center at Texas A&M University.

Burrows, Ronald Allen

1995-01-01T23:59:59.000Z

284

Less than severe worst case accidents  

Science Conference Proceedings (OSTI)

Many systems can provide tremendous benefit if operating correctly, produce only an inconvenience if they fail to operate, but have extreme consequences if they are only partially disabled such that they operate erratically or prematurely. In order to assure safety, systems are often tested against the most severe environments and accidents that are considered possible to ensure either safe operation or safe failure. However, it is often the less severe environments which result in the ``worst case accident`` since these are the conditions in which part of the system may be exposed or rendered unpredictable prior to total system failure. Some examples of less severe mechanical, thermal, and electrical environments which may actually be worst case are described as cautions for others in industries with high consequence operations or products.

Sanders, G.A.

1996-08-01T23:59:59.000Z

285

Characterization of a nuclear accident dosimeter  

Science Conference Proceedings (OSTI)

The 23rd nuclear accident dosimetry intercomparison was held during the week of June 12--16, 1995 at Los Alamos National Laboratory. This report presents the results of this event, referred to as NAD 23, as related to the performance of Sandia National Laboratories (SNL) personal nuclear accident dosimeter (PNAD). Two separate critical assemblies, SHEBA and Godiva, were used to generate seven separate neutron spectra for use in dose comparisons. SNL`s PNAD measured absorbed doses that were within +16 to +26% of the reference doses. In addition, a preliminary investigation was undertaken to determine the feasibility of using the data obtained from an irradiated PNAD to correct for body orientation. This portion of the experiment was performed with a TRIGA reactor at the Nuclear Science Center at Texas A and M University.

Burrows, R.A.

1995-12-01T23:59:59.000Z

286

US Department of Energy Chernobyl accident bibliography  

SciTech Connect

This bibliography has been prepared by Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE) Office of Health and Environmental Research to provide bibliographic information in a usable format for research studies relating to the Chernobyl nuclear accident that occurred in the Ukrainian Republic, USSR in 1986. This report is a product of the Chernobyl Database Management project. The purpose of this project is to produce and maintain an information system that is the official United States repository for information related to the accident. Two related products prepared for this project are the Chernobyl Bibliographic Search System (ChernoLit{trademark}) and the Chernobyl Radiological Measurements Information System (ChernoDat). This report supersedes the original release of Chernobyl Bibliography (Carr and Mahaffey, 1989). The original report included about 2200 references. Over 4500 references and an index of authors and editors are included in this report.

Kennedy, R.A.; Mahaffey, J.A.; Carr, F. Jr.

1992-04-01T23:59:59.000Z

287

Methods for detecting and locating leaks in containment facilities using electrical potential data and electrical resistance tomographic imaging techniques  

DOE Patents (OSTI)

Methods are provided for detecting and locating leaks in liners used as barriers in the construction of landfills, surface impoundments, water reservoirs, tanks, and the like. Electrodes are placed in the ground around the periphery of the facility, in the leak detection zone located between two liners if present, and/or within the containment facility. Electrical resistivity data is collected using these electrodes. This data is used to map the electrical resistivity distribution beneath the containment liner between two liners in a double-lined facility. In an alternative embodiment, an electrode placed within the lined facility is driven to an electrical potential with respect to another electrode placed at a distance from the lined facility (mise-a-la-masse). Voltage differences are then measured between various combinations of additional electrodes placed in the soil on the periphery of the facility, the leak detection zone, or within the facility. A leak of liquid though the liner material will result in an electrical potential distribution that can be measured at the electrodes. The leak position is located by determining the coordinates of an electrical current source pole that best fits the measured potentials with the constraints of the known or assumed resistivity distribution.

Daily, William D. (Livermore, CA); Laine, Daren L. (San Antonio, TX); Laine, Edwin F. (Alamo, CA)

1997-01-01T23:59:59.000Z

288

Methods for detecting and locating leaks in containment facilities using electrical potential data and electrical resistance tomographic imaging techniques  

DOE Patents (OSTI)

Methods are provided for detecting and locating leaks in liners used as barriers in the construction of landfills, surface impoundments, water reservoirs, tanks, and the like. Electrodes are placed in the ground around the periphery of the facility, in the leak detection zone located between two liners if present, and/or within the containment facility. Electrical resistivity data is collected using these electrodes. This data is used to map the electrical resistivity distribution beneath the containment liner between two liners in a double-lined facility. In an alternative embodiment, an electrode placed within the lined facility is driven to an electrical potential with respect to another electrode placed at a distance from the lined facility (mise-a-la-masse). Voltage differences are then measured between various combinations of additional electrodes placed in the soil on the periphery of the facility, the leak detection zone, or within the facility. A leak of liquid though the liner material will result in an electrical potential distribution that can be measured at the electrodes. The leak position is located by determining the coordinates of an electrical current source pole that best fits the measured potentials with the constraints of the known or assumed resistivity distribution. 6 figs.

Daily, W.D.; Laine, D.L.; Laine, E.F.

1997-08-26T23:59:59.000Z

289

Methods for detecting and locating leaks in containment facilities using electrical potential data and electrical resistance tomographic imaging techniques  

DOE Patents (OSTI)

Methods are provided for detecting and locating leaks in liners used as barriers in the construction of landfills, surface impoundments, water reservoirs, tanks, and the like. Electrodes are placed in the ground around the periphery of the facility, in the leak detection zone located between two liners if present, and/or within the containment facility. Electrical resistivity data is collected using these electrodes. This data is used to map the electrical resistivity distribution beneath the containment liner or between two liners in a double-lined facility. In an alternative embodiment, an electrode placed within the lined facility is driven to an electrical potential with respect to another electrode placed at a distance from the lined facility (mise-a-la-masse). Voltage differences are then measured between various combinations of additional electrodes placed in the soil on the periphery of the facility, the leak detection zone, or within the facility. A leak of liquid through the liner material will result in an electrical potential distribution that can be measured at the electrodes. The leak position is located by determining the coordinates of an electrical current source pole that best fits the measured potentials with the constraints of the known or assumed resistivity distribution.

Daily, William D. (Livermore, CA); Laine, Daren L. (San Anotonio, TX); Laine, Edwin F. (Penn Valley, CA)

2001-01-01T23:59:59.000Z

290

Helium leak testing of a radioactive contaminated vessel under high pressure in a contaminated environment  

Science Conference Proceedings (OSTI)

At ANL-W, with the shutdown of EBR-II, R&D has evolved from advanced reactor design to the safe handling, processing, packaging, and transporting spent nuclear fuel and nuclear waste. New methods of processing spent fuel rods and transforming contaminated material into acceptable waste forms are now in development. Storage of nuclear waste is a high interest item. ANL-W is participating in research of safe storage of nuclear waste, with the WIPP (Waste Isolation Pilot Plant) site in New Mexico the repository. The vessel under test simulates gas generated by contaminated materials stored underground at the WIPP site. The test vessel is 90% filled with a mixture of contaminated material and salt brine (from WIPP site) and pressurized with N2-1% He at 2500 psia. Test acceptance criteria is leakage jar method is used to determine leakage rate using a mass spectrometer leak detector (MSLD). The efficient MSLD and an Al bell jar replaced a costly, time consuming pressure decay test setup. Misinterpretation of test criterion data caused lengthy delays, resulting in the development of a unique procedure. Reevaluation of the initial intent of the test criteria resulted in leak tolerances being corrected and test efficiency improved.

Winter, M.E.

1996-10-01T23:59:59.000Z

291

Helium leak testing of a radioactive contaminated vessel under high pressure in a contaminated environment  

SciTech Connect

At ANL-W, with the shutdown of EBR-II, R&D has evolved from advanced reactor design to the safe handling, processing, packaging, and transporting spent nuclear fuel and nuclear waste. New methods of processing spent fuel rods and transforming contaminated material into acceptable waste forms are now in development. Storage of nuclear waste is a high interest item. ANL-W is participating in research of safe storage of nuclear waste, with the WIPP (Waste Isolation Pilot Plant) site in New Mexico the repository. The vessel under test simulates gas generated by contaminated materials stored underground at the WIPP site. The test vessel is 90% filled with a mixture of contaminated material and salt brine (from WIPP site) and pressurized with N2-1% He at 2500 psia. Test acceptance criteria is leakage < 10{sup -7} cc/seconds at 2500 psia. The bell jar method is used to determine leakage rate using a mass spectrometer leak detector (MSLD). The efficient MSLD and an Al bell jar replaced a costly, time consuming pressure decay test setup. Misinterpretation of test criterion data caused lengthy delays, resulting in the development of a unique procedure. Reevaluation of the initial intent of the test criteria resulted in leak tolerances being corrected and test efficiency improved.

Winter, M.E.

1996-10-01T23:59:59.000Z

292

Identifying structural damage from images  

E-Print Network (OSTI)

Photographs of damaged buildings in Bam, Iran. . . . . . . .patches of Bam, Iran. . . . . . . . . Photo pictures ofBitemporal images of an urban region of Bam, Iran. . . . .

Chen, ZhiQiang

2009-01-01T23:59:59.000Z

293

Corrosion Damage Models and Sustainability  

Science Conference Proceedings (OSTI)

Presentation Title, Corrosion Damage Models and Sustainability ... Abstract Scope, The ability of industry to make sustainable choices in the future that optimize...

294

Simple catalytic cell for restoring He leak detector sensitivity on vacuum systems with high D{sub 2} backgrounds  

SciTech Connect

The DIII{endash}D National Fusion Facility at General Atomics (GA) focuses on plasma physics and fusion energy science. The DIII{endash}D tokamak is a 35 m{sup 3} toroidal vacuum vessel with over 200 ports for diagnostic instrumentation, cryogenics, microwave heating, and four large neutral beam injectors. Maintaining vacuum in the 10{sup {minus}8}&hthinsp;Torr range is crucial for producing high performance plasma discharges. He leak checking the DIII{endash}D tokamak and the neutral beamlines has historically been difficult. D{sub 2} is used as the fuel gas in most plasma discharges and neutral beams. After plasma operations, D{sub 2} outgassing from the torus walls and internal beamline components can exceed 10{sup {minus}4}&hthinsp;std&hthinsp;cm{sup 3}/s. The mass of the D{sub 2} molecule (4.028 u) is indistinguishable from that of the He atom (4.003 u) to a standard mass spectrometer leak detector. High levels of D{sub 2} reduce leak detector sensitivity and effectively mask the He trace gas signal rendering normal leak checking techniques ineffective. A simple apparatus was developed at GA to address these problems. It consists of a palladium based catalyst cell and associated valves and piping placed in series with the leak detector. This reduces the D{sub 2} throughput by a factor greater than 10&hthinsp;000, restoring leak detector sensitivity. This article will briefly discuss the development of the cell, the physical processes involved, the tests performed to quantify and optimize the processes, and the operational results at DIII{endash}D. {copyright} {ital 1999 American Vacuum Society.}

Busath, J.; Chiu, H.K. [General Atomics, San Diego, California 92186-5608 (United States)

1999-07-01T23:59:59.000Z

295

Simple catalytic cell for restoring He leak detector sensitivity on vacuum systems with high D{sub 2} backgrounds  

SciTech Connect

The DIII-D National Fusion Facility at General Atomics focuses on plasma physics and fusion energy science. The DIII-D tokamak is a 35 m{sup 3} toroidal vacuum vessel with over 200 ports for diagnostic instrumentation, cryogenics, microwave heating, and four large neutral beam injectors. Maintaining vacuum in the 10{sup {minus}8} Torr range is crucial for producing high performance plasma discharges. He leak checking the DIII-D tokamak and the neutral beamlines has historically been difficult. D{sub 2} is used as the fuel gas in most plasma discharges and neutral beams. After plasma operations, D{sub 2} out-gassing from the torus walls and internal beamline components can exceed 10{sup {minus}4} std cc/s. The mass of the D{sub 2} molecule (4.028 u) is indistinguishable from that of the He atom (4.003 u) to a standard mass spectrometer leak detector. High levels of D{sub 2} reduce leak detector sensitivity and effectively mask the He trace gas signal rendering normal leak checking techniques ineffective. A simple apparatus was developed at GA to address these problems. It consists of a palladium based catalyst cell and associated valves and piping placed in series with the leak detector. This reduces the D{sub 2} throughput by a factor greater than 10,000, restoring leak detector sensitivity. This paper will briefly discuss the development of the cell, the physical processes involved, the tests performed to quantify and optimize the processes, and the operational results at DIII-D.

Busath, J.; Chiu, H.K.

1998-12-01T23:59:59.000Z

296

A framework for the assessment of severe accident management strategies  

SciTech Connect

Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.

Kastenberg, W.E. [ed.; Apostolakis, G.; Dhir, V.K. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering] [and others

1993-09-01T23:59:59.000Z

297

TMI-2 accident: core heat-up analysis  

SciTech Connect

This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions.

Ardron, K.H.; Cain, D.G.

1981-01-01T23:59:59.000Z

298

Preliminary dose assessment of the Chernobyl accident  

Science Conference Proceedings (OSTI)

From the major accident at Unit 4 of the Chernobyl nuclear power station, a plume of airborne radioactive fission products was initially carried northwesterly toward Poland, thence toward Scandinavia and into Central Europe. Reports of the levels of radioactivity in a variety of media and of external radiation levels were collected in the Department of Energy's Emergency Operations Center and compiled into a data bank. Portions of these and other data which were obtained directly from published and official reports were utilized to make a preliminary assessment of the extent and magnitude of the external dose to individuals downwind from Chernobyl. Radioactive /sup 131/I was the predominant fission product. The time of arrival of the plume and the maximum concentrations of /sup 131/I in air, vegetation and milk and the maximum reported depositions and external radiation levels have been tabulated country by country. A large amount of the total activity in the release was apparently carried to a significant elevation. The data suggest that in areas where rainfall occurred, deposition levels were from ten to one-hundred times those observed in nearby ''dry'' locations. Sufficient spectral data were obtained to establish average release fractions and to establish a reference spectra of the other nuclides in the release. Preliminary calculations indicated that the collective dose equivalent to the population in Scandinavia and Central Europe during the first year after the Chernobyl accident would be about 8 x 10/sup 6/ person-rem. From the Soviet report, it appears that a first year population dose of about 2 x 10/sup 7/ person-rem (2 x 10/sup 5/ Sv) will be received by the population who were downwind of Chernobyl within the U.S.S.R. during the accident and its subsequent releases over the following week. 32 refs., 14 figs., 20 tabs.

Hull, A.P.

1987-01-01T23:59:59.000Z

299

Sec. Herrington Leads Delegation in Response to Chernobyl Accident...  

NLE Websites -- All DOE Office Websites (Extended Search)

Sec. Herrington Leads Delegation in Response to Chernobyl Accident | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering...

300

Next-generation nuclear fuel withstands high-temperature accident...  

NLE Websites -- All DOE Office Websites (Extended Search)

(more than 200 degrees Celsius greater than postulated accident conditions) most fission products remained inside the fuel particles, which each boast their own primary...

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

HEALTH EFFECTS OF THE NUCLEAR ACCIDENT AT THREE MILE ISLAND  

E-Print Network (OSTI)

In) Symposium on Nuclear Reactor Safety: Perspective. Ahealth effects of the nuclear reactor accident at Three Mile50-mile radius of the nuclear reactor site, approximately

Fabrikant, J.I.

2010-01-01T23:59:59.000Z

302

Median Light Rail Crossing: Accident Causation And Countermeasures  

E-Print Network (OSTI)

Integration of Light Rail Transit Into City Streets. TCRPInfluencing Safety at Highway-Rail Grade Crossings. InK. , W. Hucke and W. Berg. Rail Highway Crossing Accident

Coifman, Benjamin; Bertini, Robert L.

1997-01-01T23:59:59.000Z

303

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident  

Science Conference Proceedings (OSTI)

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

Su'ud, Zaki; Anshari, Rio [Nuclear and Biophysics Research Group, Dept. of Physics, Bandung Institute of Technology, Jl.Ganesha 10, Bandung, 40132 (Indonesia)

2012-06-06T23:59:59.000Z

304

The Pipeline Still Leaks and More Than You Think: A Status Report on Gender Diversity in Biomedical Engineering  

E-Print Network (OSTI)

The Pipeline Still Leaks and More Than You Think: A Status Report on Gender Diversity in Biomedical is evidence of a still leaky pipeline in our discipline. In addition, the percentage of women faculty members in the pipeline--are reviewed. Keywords--Women, Engineering, Barriers, Bias. INTRODUCTION The lack of diversity

Bhatia, Sangeeta

305

A novel neural model-based approach to leak detection and localization in oil pipelines for environmental protection  

Science Conference Proceedings (OSTI)

Monitoring oil transporting pipelines is an important task for economical and safe operation, loss prevention, and environmental protection from crude oil emission. The leak detection of oil pipelines, therefore, plays a key role in the overall integrity ... Keywords: environmental and safety systems, fault and uncertainty modeling in dynamical systems, neural nets, process supervision

Alireza Paivar; Karim Salahshoor; Farzad Hourfar

2006-11-01T23:59:59.000Z

306

Investigation of leaks in fiberglass-reinforced pressure vessels by direct observation of hollow fibers in glass cloth  

SciTech Connect

A simple method of visual observation of hollow fibers within fiberglass cloth has been developed. This visualization can aid in determining the contribution these fibers make toward leaks observed in fiberglass-reinforced epoxy resin pressure or vacuum vessels. Photographs and frequency data of these hollow fibers are provided. 3 figs.

McAdams, J.

1988-01-01T23:59:59.000Z

307

Nuclear fuel cycle facility accident analysis handbook  

Science Conference Proceedings (OSTI)

The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

NONE

1998-03-01T23:59:59.000Z

308

Excitation optimization for damage detection  

SciTech Connect

A technique is developed to answer the important question: 'Given limited system response measurements and ever-present physical limits on the level of excitation, what excitation should be provided to a system to make damage most detectable?' Specifically, a method is presented for optimizing excitations that maximize the sensitivity of output measurements to perturbations in damage-related parameters estimated with an extended Kalman filter. This optimization is carried out in a computationally efficient manner using adjoint-based optimization and causes the innovations term in the extended Kalman filter to be larger in the presence of estimation errors, which leads to a better estimate of the damage-related parameters in question. The technique is demonstrated numerically on a nonlinear 2 DOF system, where a significant improvement in the damage-related parameter estimation is observed.

Bement, Matthew T [Los Alamos National Laboratory; Bewley, Thomas R [UCSD

2009-01-01T23:59:59.000Z

309

Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building  

DOE Green Energy (OSTI)

About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures.

Alvares, N.J.; Beason, D.G.; Eidem, G.R.

1982-06-01T23:59:59.000Z

310

DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Video User' s Guide Video User' s Guide DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING RADIOACTIVE MATERIAL DISCLAIMER Viewing this video and completing the enclosed printed study material do not by themselves provide sufficient skills to safely engage in or perform duties related to emergency response to a transportation accident involving radioactive material. Meeting that goal is beyond

311

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation  

Science Conference Proceedings (OSTI)

This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

CROWE, R.D.; PIEPHO, M.G.

2000-03-23T23:59:59.000Z

312

Variable selection and ranking for analyzing automobile traffic accident data  

Science Conference Proceedings (OSTI)

Variable ranking and feature selection are important concepts in data mining and machine learning. This paper introduces a new variable ranking technique named Sum Max Gain Ratio (SMGR). The new technique is evaluated within the domain of traffic accident ... Keywords: decision tree, traffic accident data, variable and feature selection, variable ranking

Huanjing Wang; Allen Parrish; Randy K. Smith; Susan Vrbsky

2005-03-01T23:59:59.000Z

313

Assessment of Existing Plant Instrumentation for Severe Accident Management  

Science Conference Proceedings (OSTI)

During an accident, information would be needed for diagnosing a plant's status and confirming its response to mitigative actions. It is important to determine the information necessary for severe accident management and to ensure that this information could be derived from plant instrumentation.

1993-12-01T23:59:59.000Z

314

A review of the Melcor Accident Consequence Code System (MACCS): Capabilities and applications  

Science Conference Proceedings (OSTI)

MACCS was developed at Sandia National Laboratories (SNL) under U.S. Nuclear Regulatory Commission (NRC) sponsorship to estimate the offsite consequences of potential severe accidents at nuclear power plants (NPPs). MACCS was publicly released in 1990. MACCS was developed to support the NRC`s probabilistic safety assessment (PSA) efforts. PSA techniques can provide a measure of the risk of reactor operation. PSAs are generally divided into three levels. Level one efforts identify potential plant damage states that lead to core damage and the associated probabilities, level two models damage progression and containment strength for establishing fission-product release categories, and level three efforts evaluate potential off-site consequences of radiological releases and the probabilities associated with the consequences. MACCS was designed as a tool for level three PSA analysis. MACCS performs probabilistic health and economic consequence assessments of hypothetical accidental releases of radioactive material from NPPs. MACCS includes models for atmospheric dispersion and transport, wet and dry deposition, the probabilistic treatment of meteorology, environmental transfer, countermeasure strategies, dosimetry, health effects, and economic impacts. The computer systems MACCS is designed to run on are the 386/486 PC, VAX/VMS, E3M RISC S/6000, Sun SPARC, and Cray UNICOS. This paper provides an overview of MACCS, reviews some of the applications of MACCS, international collaborations which have involved MACCS, current developmental efforts, and future directions.

Young, M.

1995-02-01T23:59:59.000Z

315

Web Based Course: SAF-230DE, Accident Investigation Overview Promotional Video  

Energy.gov (U.S. Department of Energy (DOE))

This course that provides an overview of the fundamentals of accident investigation. The course is intended to meet the every five year refresher training requirement for DOE Federal Accident Investigators under DOE O 225.1B, Accident Investigations.

316

Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices  

SciTech Connect

This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs.

Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L. (Science Applications International Corp., Albuquerque, NM (USA); Sandia National Labs., Albuquerque, NM (USA))

1989-08-01T23:59:59.000Z

317

Naval Spent Fuel Rail Shipment Accident Exercise Objectives  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NAVAL SPENT FUEL RAIL SHIPMENT NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the Naval spent fuel shipping container characteristics and shipping practices * Gain understanding of how the NNPP escorts who accompany the spent fuel shipments will interact with civilian emergency services representatives g y p * Allow civilian emergency services agencies the opportunity to evaluate their response to a pp y p simulated accident * Gain understanding of how the communications links that would be activated in an accident involving a Naval spent fuel shipment would work 1 NTSF May 11 ACCIDENT EXERCISE TYPICAL TIMELINE * Conceptual/Organizational Meeting - April 6 E R T i d it t t d TYPICAL TIMELINE

318

MELCOR accident analysis for ARIES-ACT  

Science Conference Proceedings (OSTI)

We model a loss of flow accident (LOFA) in the ARIES-ACT1 tokamak design. ARIES-ACT1 features an advanced SiC blanket with LiPb as coolant and breeder, a helium cooled steel structural ring and tungsten divertors, a thin-walled, helium cooled vacuum vessel, and a room temperature water-cooled shield outside the vacuum vessel. The water heat transfer system is designed to remove heat by natural circulation during a LOFA. The MELCOR model uses time-dependent decay heats for each component determined by 1-D modeling. The MELCOR model shows that, despite periodic boiling of the water coolant, that structures are kept adequately cool by the passive safety system.

Paul W. Humrickhouse; Brad J. Merrill

2012-08-01T23:59:59.000Z

319

Angular dependence of a simple accident dosimeter  

SciTech Connect

A simple dosimeter made of a sulfur tablet, bare and cadmium covered indium foils and a cadmium covered copper foil has been modeled using MCNP5. Studies of the model without phantom or other confounding factors have shown that the cross sections and fluence-to-dose factors generated by the Monte Carlo method agree with those generated by analytic expressions for the high energy component. The threshold cross sections for the detectors on a phantom were calculated. The resulting doses assigned agree well with exposures made to three critical assemblies. In this study the angular dependence on a phantom is studied and compared with measurements taken on the GODIVA reactor. The dosimeter positions on the phantom are facing the source, on the back and the side. In previous papers the modeling of a simple dosimeter made of a sulfur tablet, bare and cadmium covered indium foils and a cadmium covered copper foil has been modeled using MCNP5. The conclusion made was that most of the neutron dose from criticality assemblies results from the high energy neutron fluences determined by the sulfur and indium detectors. The results using doses measured from the GODIVA, SHEBA, and bare and lead shielded SILENE reactors confirmed this. The angular dependence of an accident dosemeter is of interest in evaluating the exposure of personnel. To investigate this effect accident dosemeters were placed on a phantom and exposed to the GODIVA reactor at phantom orientations of 0{sup o}, 45{sup o}, 90{sup o}, 135{sup o}, and 180{sup o} to the assembly center line.

Devine, R. T. (Robert T.); Romero, L. L. (Leonard L.); Olsher, R. H. (Richard H.)

2004-01-01T23:59:59.000Z

320

DETECTION OF HISTORICAL PIPELINE LEAK PLUMES USING NON-INTRUSIVE SURFACE-BASED GEOPHYSICAL TECHNIQUES AT THE HANFORD NUCLEAR SITE WASHINGTON USA  

Science Conference Proceedings (OSTI)

Historical records from the Department of Energy Hanford Nuclear Reservation (in eastern WA) indicate that ruptures in buried waste transfer pipelines were common between the 1940s and 1980s, which resulted in unplanned releases (UPRs) of tank: waste at numerous locations. A number of methods are commercially available for the detection of active or recent leaks, however, there are no methods available for the detection of leaks that occurred many years ago. Over the decades, leaks from the Hanford pipelines were detected by visual observation of fluid on the surface, mass balance calculations (where flow volumes were monitored), and incidental encounters with waste during excavation or drilling. Since these detection methods for historic leaks are so limited in resolution and effectiveness, it is likely that a significant number of pipeline leaks have not been detected. Therefore, a technology was needed to detect the specific location of unknown pipeline leaks so that characterization technologies can be used to identify any risks to groundwater caused by waste released into the vadose zone. A proof-of-concept electromagnetic geophysical survey was conducted at an UPR in order to image a historical leak from a waste transfer pipeline. The survey was designed to test an innovative electromagnetic geophysical technique that could be used to rapidly map the extent of historical leaks from pipelines within the Hanford Site complex. This proof-of-concept test included comprehensive testing and analysis of the transient electromagnetic method (TEM) and made use of supporting and confirmatory geophysical methods including ground penetrating radar, magnetics, and electrical resistivity characterization (ERC). The results for this initial proof-of-concept test were successful and greatly exceeded the expectations of the project team by providing excellent discrimination of soils contaminated with leaked waste despite the interference from an electrically conductive pipe.

SKORSKA MB; FINK JB; RUCKER DF; LEVITT MT

2010-12-02T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

In-vessel Zircaloy oxidation/hydrogen generation behavior during severe accidents  

DOE Green Energy (OSTI)

In-vessel Zircaloy oxidation and hydrogen generation data from various US Nuclear Regulatory Commission severe-fuel damage test programs are presented and compared, where the effects of Zircaloy melting, bundle reconfiguration, and bundle quenching by reflooding are assessed for common findings. The experiments evaluated include fuel bundles incorporating fresh and previously irradiated fuel rods, as well as control rods. Findings indicate that the extent of bundle oxidation is largely controlled by steam supply conditions and that high rates of hydrogen generation continued after melt formation and relocation. Likewise, no retardation of hydrogen generation was noted for experiments which incorporated control rods. Metallographic findings indicate extensive oxidation of once-molten Zircaloy bearing test debris. Such test results indicate no apparent limitations to Zircaloy oxidation for fuel bundles subjected to severe-accident coolant-boiloff conditions. 46 refs., 22 figs., 12 tabs.

Cronenberg, A.W. (Science and Engineering Associates, Inc., Albuquerque, NM (USA))

1990-09-01T23:59:59.000Z

322

Criteria for calculating the efficiency of HEPA filters during and after design basis accidents  

SciTech Connect

We have reviewed the literature on the performance of high efficiency particulate air (HEPA) filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be structurally damaged and have a residual efficiency of 0%. Despite the many studies on HEPA filter performance under adverse conditions, there are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen when there was insufficient data.

Bergman, W. [Lawrence Livermore National Lab., CA (United States); First, M.W. [Harvard School of Public Health, Boston, MA (United States); Anderson, W.L. [Consultant, LaPlata, MD (United States); Gilbert, H. [Consultant, McLean, VA (United States); Jacox, J.W. [Consultant, Columbus, OH (United States)

1994-12-01T23:59:59.000Z

323

Review of the Oconee-3 probabilistic risk assessment: external events, core damage frequency. Volume 2  

SciTech Connect

A review of the Oconee-3 Probabilistic Risk Assessment (OPRA) was conducted with the broad objective of evaluating qualitatively and quantitatively (as much as possible) the OPRA assessment of the important sequences that are ''externally'' generated and lead to core damage. The review included a technical assessment of the assumptions and methods used in the OPRA within its stated objective and with the limited information available. Within this scope, BNL performed a detailed reevaluation of the accident sequences generated by internal floods and earthquakes and a less detailed review (in some cases a scoping review) for the accident sequences generated by fires, tornadoes, external floods, and aircraft impact. 12 refs., 24 figs., 31 tabs.

Hanan, N.A.; Ilberg, D.; Xue, D.; Youngblood, R.; Reed, J.W.; McCann, M.; Talwani, T.; Wreathall, J.; Kurth, P.D.; Bandyopadhyay, K.

1986-03-01T23:59:59.000Z

324

Microstructural Characterization of Damage Mechanisms of Graphite ...  

Science Conference Proceedings (OSTI)

Symposium, Nanostructured Materials for Lithium Ion Batteries and for Supercapacitors. Presentation Title, Microstructural Characterization of Damage ...

325

AIRBORNE, OPTICAL REMOTE SENSING OF METHANE AND ETHANE FOR NATURAL GAS PIPELINE LEAK DETECTION  

SciTech Connect

Ophir Corporation was awarded a contract by the U. S. Department of Energy, National Energy Technology Laboratory under the Project Title ''Airborne, Optical Remote Sensing of Methane and Ethane for Natural Gas Pipeline Leak Detection'' on October 14, 2002. This second six-month technical report summarizes the progress made towards defining, designing, and developing the hardware and software segments of the airborne, optical remote methane and ethane sensor. The most challenging task to date has been to identify a vendor capable of designing and developing a light source with the appropriate output wavelength and power. This report will document the work that has been done to identify design requirements, and potential vendors for the light source. Significant progress has also been made in characterizing the amount of light return available from a remote target at various distances from the light source. A great deal of time has been spent conducting laboratory and long-optical path target reflectance measurements. This is important since it helps to establish the overall optical output requirements for the sensor. It also reduces the relative uncertainty and risk associated with developing a custom light source. The data gathered from the optical path testing has been translated to the airborne transceiver design in such areas as: fiber coupling, optical detector selection, gas filters, and software analysis. Ophir will next, summarize the design progress of the transceiver hardware and software development. Finally, Ophir will discuss remaining project issues that may impact the success of the project.

Jerry Myers

2003-11-12T23:59:59.000Z

326

The depth of the oil/brine interface and crude oil leaks in SPR caverns  

Science Conference Proceedings (OSTI)

Monitoring wellhead pressure evolution is the best method of detecting crude oil leaks in SPR caverns while oil/brine interface depth measurements provide additional insight. However, to fully utilize the information provided by these interface depth measurements, a thorough understanding of how the interface movement corresponds to cavern phenomena, such as salt creep, crude oil leakage, and temperature equilibration, as well as to wellhead pressure, is required. The time evolution of the oil/brine interface depth is a function of several opposing factors. Cavern closure due to salt creep and crude oil leakage, if present, move the interface upward. Brine removal and temperature equilibration of the oil/brine system move the interface downward. Therefore, the relative magnitudes of these factors determine the net direction of interface movement. Using a mass balance on the cavern fluids, coupled with a simplified salt creep model for closure in SPR caverns, the movement of the oil/brine interface has been predicted for varying cavern configurations, including both right-cylindrical and carrot-shaped caverns. Three different cavern depths and operating pressures have been investigated. In addition, the caverns were investigated at four different points in time, allowing for varying extents of temperature equilibration. Time dependent interface depth changes of a few inches to a few feet were found to be characteristic of the range of cases studied. 5 refs, 19 figs., 1 tab.

Heffelfinger, G.S.

1991-06-01T23:59:59.000Z

327

The world oil market and OPEC behavior: The leak-producer price leader model  

SciTech Connect

This is an economic study of the world's oil market in which OPEC plays the central role in determining the oil supply and price. Understanding OPEC's behavior is at the core of understanding the world's oil market. However, oil is a resource belonging to the family of natural resources known as exhaustible. We do not produce oil; we only extract and distribute a fixed amount of the resource over generations. Optimal extraction is a matter of concern to both suppliers and consumers. First, it is shown that using the traditional theory of producers behavior in the conventional commodity markets to explain extractors behavior in exhaustible resource markets is completely wrong. Second, current models of OPEC behavior are reviewed. Third, an alternative model is introduced. Previous authors have not directed their models to give explanations to the peculiar observations in oil market. This model divides the world's oil suppliers into: the free riders (non-OPEC oil producers), the OPEC hawks (a group within OPEC) and the leak-producer price leader (Saudi Arabia). Three factors, namely relatively big oil reserves, no other sources of income, and the avoidance of the so-called backstop technology make Saudi Arabia more interested in lower oil prices than are other oil extractors.

Aboalela, A.A.

1988-01-01T23:59:59.000Z

328

METHOD AND APPARATUS FOR THE DETECTION OF LEAKS IN PIPE LINES  

DOE Patents (OSTI)

A method is described for detecting leaks in pipe lines carrying fluid. The steps include the following: injecting a radioactive solution into a fluid flowing in the line; flushing the line clear of the radioactive solution; introducing a detector-recorder unit, comprising a radioactivity radiation detector and a recorder which records the detector signal over a time period at a substantially constant speed, into the line in association with a go-devil capable of propelling the detector-recorder unit through the line in the direction of the fluid flow at a substantia1ly constant velocity; placing a series of sources of radioactivity at predetermined distances along the downstream part of the line to make a characteristic signal on the recorder record at intervals corresponding to the location of said sources; recovering the detector-recorder unit at a downstream point along the line; transcribing the recorder record of any radioactivity detected during the travel of the detector- recorder unit in terms of distance along the line. (AEC)

Jefferson, S.; Cameron, J.F.

1961-11-28T23:59:59.000Z

329

Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components  

SciTech Connect

Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

Finn, P.A.

1991-03-01T23:59:59.000Z

330

AIRBORNE, OPTICAL REMOTE SENSING OF METHANE AND ETHANE FOR NATURAL GAS PIPLINE LEAK DETECTION  

SciTech Connect

Ophir Corporation was awarded a contract by the U. S. Department of Energy, National Energy Technology Laboratory under the Project Title ''Airborne, Optical Remote Sensing of Methane and Ethane for Natural Gas Pipeline Leak Detection'' on October 14, 2002. The third six-month technical report contains a summary of the progress made towards finalizing the design and assembling the airborne, remote methane and ethane sensor. The vendor has been chosen and is on contract to develop the light source with the appropriate linewidth and spectral shape to best utilize the Ophir gas correlation software. Ophir has expanded upon the target reflectance testing begun in the previous performance period by replacing the experimental receiving optics with the proposed airborne large aperture telescope, which is theoretically capable of capturing many times more signal return. The data gathered from these tests has shown the importance of optimizing the fiber optic receiving fiber to the receiving optic and has helped Ophir to optimize the design of the gas cells and narrowband optical filters. Finally, Ophir will discuss remaining project issues that may impact the success of the project.

Jerry Myers

2004-05-12T23:59:59.000Z

331

Investigation of Strategies for Mitigating Radiological Releases in Severe Accidents  

Science Conference Proceedings (OSTI)

The Fukushima Dai-ichi accident highlights the need to reduce the magnitude of radioactive fission product releases from BWR Mark I and II containments following beyond-design-basis events. There is no evidence that this accident has a long-term effect on public health and safety; however, the Fukushima Dai-ichi accident did result in widespread contamination of surrounding areas, both on-site and off-site. This report assesses various strategies that can be used to maintain BWR Mark I and II ...

2012-09-24T23:59:59.000Z

332

Severe accident sequences analyzed for a two-loop PWR  

Science Conference Proceedings (OSTI)

Different severe accident sequences have been analyzed for a two-loop Westinghouse pressurized water reactor (PWR) using the MELCOR code, version 1.8.4. The purpose of this study was to calculate source terms and the timing of events for severe accident sequences at this type of PWR to be used in the HAS-CAL code .The results calculated by MELCOR have been compared to results from the individual plant examination (IPE) of the Kewaunee nuclear power plant, also a two-loop Westinghouse PWR. The results of the Kewaunee IPE were obtained with the severe accident code MAAP.

Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

1997-12-01T23:59:59.000Z

333

Measured leak rates of the temporary seals in DWPF canistered waste forms after three years of on site storage  

SciTech Connect

In the summer of 1990 a study was carried out to determine the-internal pressure, relative humidity, and chemical composition of the gas within the free volume of four canistered waste forms produced at TNX in May of 1988. Three of these canistered waste forms were sealed only by temporary seals and subsequently stored in the TNX boneyard' with no protection. The fourth canister was sealed by upset resistance welding. All three canisters with temporary seals were decontaminated by aqueous frit blasting. It was important to remeasure the leak rates of these seals to ensure that leaktightness had not deteriorated during canister handling and storage prior to the time the experiment were performed. This paper details the results of two separate measurements of the leak rates of these seals.

Harbour, J.R.; Miller, T.J.

1992-04-06T23:59:59.000Z

334

Measured leak rates of the temporary seals in DWPF canistered waste forms after three years of on site storage  

SciTech Connect

In the summer of 1990 a study was carried out to determine the-internal pressure, relative humidity, and chemical composition of the gas within the free volume of four canistered waste forms produced at TNX in May of 1988. Three of these canistered waste forms were sealed only by temporary seals and subsequently stored in the TNX `boneyard` with no protection. The fourth canister was sealed by upset resistance welding. All three canisters with temporary seals were decontaminated by aqueous frit blasting. It was important to remeasure the leak rates of these seals to ensure that leaktightness had not deteriorated during canister handling and storage prior to the time the experiment were performed. This paper details the results of two separate measurements of the leak rates of these seals.

Harbour, J.R.; Miller, T.J.

1992-04-06T23:59:59.000Z

335

Computational Assessment of the GT-MHR Graphite Core Support Structural Integrity in Air-Ingress Accident Condition  

Science Conference Proceedings (OSTI)

The objective of this project was to perform stress analysis for graphite support structures of the General Atomics 600 MWth GT-MHR prismatic core design using ABAQUS (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR was analyzed based on the change of temperature, burn-off and corrosion depth during the accident period predicted by GAMMA, a multi-dimensional gas multi-component mixture analysis code developed in the Republic of Korea (ROK)/United States (US) International Nuclear Engineering Research Initiative (I-NERI) project. Both the loading and thermal stresses were analyzed, but the thermal stress was not significant, leaving the loading stress to be the major factor. The mechanical strengths are exceeded between 11 to 11.5 days after loss-of-coolant-accident (LOCA), corresponding to 5.5 to 6 days after the start of natural convection.

Jong B. Lim; Eung S. Kim; Chang H. Oh; Richard R. Schultz; David A. Petti

2008-10-01T23:59:59.000Z

336

Experimental results from containment piping bellows subjected to severe accident conditions: Results from bellows tested in corroded conditions. Volume 2  

Science Conference Proceedings (OSTI)

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of nineteen bellows have been tested. Thirteen bellows were tested in ``like-new`` condition (results reported in Volume 1), and six were tested in a corroded condition. The tests showed that bellows in ``like-new`` condition are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage, while those in a corroded condition did not perform as well, depending on the amount of corrosion. The corroded bellows test program and results are presented in this report.

Lambert, L.D.; Parks, M.B. [Sandia National Labs., Albuquerque, NM (United States)

1995-10-01T23:59:59.000Z

337

Field Guide: Bearing Damage Mechanisms  

Science Conference Proceedings (OSTI)

Electric Power Research Institute (EPRI) report 1021780, Manual of Bearing Failures and Repair in Power Plant Rotating Equipment, 2011 Update, is a comprehensive document on the subject of fluid film bearing damage modes. This field guide provides a pocket reference based upon the content of that report. ...

2012-11-06T23:59:59.000Z

338

Emergency Response to a Transportation Accident Involving Radioactive  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Response to a Transportation Accident Involving Response to a Transportation Accident Involving Radioactive Material Emergency Response to a Transportation Accident Involving Radioactive Material The purpose of this User's Guide is to provide instructors with an overview of the key points covered in the video. The Student Handout portion of this Guide is designed to assist the instructor in reviewing those points with students. The Student Handout should be distributed to students after the video is shown and the instructor should use the Guide to facilitate a discussion on each response disciplines' activities or duties at the scene. During this discussion, the instructor can present response scenarios, each of which would have a different discipline arriving first at the accident scene. The purpose of this discussion

339

Failsafe : living with man-made disaster and accident  

E-Print Network (OSTI)

"There is no progress with out progress of the catastrophe." Virilio. This thesis project proposes that technological solutions in the design of our systems are not enough to prevent 'man-made' accident. Social, organisational ...

Higgins, Saoirse, 1966-

2004-01-01T23:59:59.000Z

340

Environment/Health/Safety/Security (EHSS): Report an Accident...  

NLE Websites -- All DOE Office Websites (Extended Search)

Report an Accident or Incident car and foot The law and DOE require prompt notification of all work-related EHS incidentsaccidents. Report all such events immediately to your...

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

HEALTH EFFECTS OF THE NUCLEAR ACCIDENT AT THREE MILE ISLAND  

E-Print Network (OSTI)

within 50 miles of the nuclear power plant was estimated tothe radiation from the nuclear power plant accident. From anand the Peach Bottom nuclear power plants, like the general

Fabrikant, J.I.

2010-01-01T23:59:59.000Z

342

Accidents, engineering and history at NASA: 1967-2003  

E-Print Network (OSTI)

The manned spaceflight program of the National Aeronautics and Space Administration (NASA) has suffered three fatal accidents: one in the Apollo program and two in the Space Transportation System (the Shuttle). These were ...

Brown, Alexander F. G. (Alexander Frederic Garder), 1970-

2009-01-01T23:59:59.000Z

343

FAQ 30-Have there been accidents involving uranium hexafluoride...  

NLE Websites -- All DOE Office Websites (Extended Search)

UF6 was released, which reacted with steam from the process and created HF and uranyl fluoride. This accident resulted in two deaths from HF inhalation and three individuals...

344

Structural evaluation of electrosleeved tubes under severe accident transients.  

Science Conference Proceedings (OSTI)

A flow stress model was developed for predicting failure of Electrosleeved PWR steam generator tubing under severe accident transients. The Electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400 C during severe accidents because of grain growth. A grain growth model and the Hall-Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data as well as high temperature failure tests on notched Electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of Electrosleeved tubes with axial cracks in the parent tube during postulated severe accident transients.

Majumdar, S.

1999-11-12T23:59:59.000Z

345

Advanced Steels for Accident Tolerant Fuel Cladding in Commercial ...  

Science Conference Proceedings (OSTI)

... (depending on the LWR system and accident scenario) while maintaining or ... Analysis of the Fragmentation of AlON and Three MgAl2O4 Spinels under...

346

Geometry features measurement of traffic accident for reconstruction based on close-range photogrammetry  

Science Conference Proceedings (OSTI)

This paper studies the feasibility of investigating a traffic accident and offering initial data for traffic accident reconstruction (TAR) using a photogrammetric technique. Compared with the conventional roller tape applied by the traffic police of ... Keywords: Accident reconstruction, Close-range photogrammetry, Direct linear transformation, Traffic accident scene, Vehicle deformation

Xinguang Du; Xianlong Jin; Xiaoyun Zhang; Jie Shen; Xinyi Hou

2009-07-01T23:59:59.000Z

347

Fallen conductor accidents: The challenge to improve safety  

SciTech Connect

What is the worst nightmare of an electric utility manager or engineer Many respond that it is an electrocution resulting from a fallen conductor accident. Few subjects in the operation of an electric utility are more emotional and sobering than this. Traditionally, a utility could do little to prevent such accidents, but some answers from research are emerging, calling for a new look at this old problem.

Aucoin, B.M.; Russell, B.D.

1992-02-01T23:59:59.000Z

348

Evaluation of Accident Frequencies at the Canister Storage Bldg (CSB)  

DOE Green Energy (OSTI)

By using the fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The evaluation demonstrates that due to low frequency of occurrences, the following design basis accidents are considered not credible (annual frequency of less than 10{sup -6}): Rearrangement of multidster overpack (MCO) internals; Gaseous release from the MCO; MCO internal hydrogen explosion; MCO external hydrogen explosion; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

LIU, Y.J.

1999-09-02T23:59:59.000Z

349

Modular Accident Analysis Program (MAAP5) Applications Assessment  

Science Conference Proceedings (OSTI)

The Modular Accident Analysis Program (MAAP) is widely used throughout North America, Europe, and the Far East to analyze plant responses over a broad spectrum of potential accident conditions. The use of MAAP continues to increase because its representation of integral plant response and short run times make this program ideal for supporting engineering evaluations. With greater use, however, the level of detail to be represented within the reactor core, reactor coolant system (RCS), and containment has...

2005-12-08T23:59:59.000Z

350

Study on drywell cooler applicability to severe accident management  

SciTech Connect

This paper concerns applicability of drywell cooler (DWC) heat removal under severe accident condition in BWR plants. Newly developed heat removal models based on DWC heat removal experiments were built into the MAAP3 code. And then, two types of Japanese BWR were selected to evaluate DWC heat removal performance under typical severe accident scenarios. According to the results of the evaluation, DWC delays or prevents containment failure or venting. (authors)

Nakagawa, Takahiro [Information and manufacturing systems division, Toshiba Plant Systems and Services Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Akinaga, Makoto [Power and Industrial Systems R and D Center, Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki, 210-0862 (Japan); Hamazaki, Ryoichi [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Matsuo, Toshihiro [Nuclear Power Engineering Department, Tokyo Electric Power Company, 1-3 Uchisaiwai-cho 1-chome, Chiyoda-ku, Tokyo 100-0011 (Japan); Hashimoto, Kouji [Nuclear Plant Engineering Department, HITACHI, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-8511 (Japan)

2004-07-01T23:59:59.000Z

351

Trees as Filters of Radioactive Fallout from the Chernobyl Accident  

E-Print Network (OSTI)

This paper is a copy of an unpublished study of the filtering effect of red maple trees (acer rubrum) on fission product fallout near Binghamton, NY, USA following the 1986 Chernobyl accident. The conclusions of this work may offer some insight into what is happening in the forests exposed to fallout from the Fukushima Daiichi Nuclear Plant accident. This posting is in memory of Noel K. Yeh.

Brownridge, James D

2011-01-01T23:59:59.000Z

352

Evaluation of accident frequencies at the canister storage building  

DOE Green Energy (OSTI)

By using the fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The evaluation demonstrates that due to low frequency of occurrences, the following design basis accidents are considered not credible (annual frequency of less than 10{sup -6}): Rearrangement of multi-canister overpack (MCO) internals; Gaseous release from the MCO; MCO internal hydrogen explosion; MCO external hydrogen explosion; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

LIU, Y.J.

1999-05-13T23:59:59.000Z

353

Review of cladding-coolant interactions during LWR accident transients  

Science Conference Proceedings (OSTI)

Some of the coolant-cladding interactions that can take place during the design basis loss-of-coolant accident and the Three Mile Island loss-of-coolant accident are analyzed. The physical manifestations of the interactions are quite similar, but the time sequences involved can cause very different end results. These results are described and a listing is given of the main research programs that are involved in coolant-cladding interaction research.

Hobson, D.O.

1980-01-01T23:59:59.000Z

354

Preliminary results of the PWR low power and shutdown accident frequencies program: Coarse screening analysis for Surry  

Science Conference Proceedings (OSTI)

This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was performed by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES). This coarse screening analysis was performed in support of the NRC staff's follow-up actions subsequent to the March 20, 1990 Vogtle incident with the objective of providing high-level qualitative insights within a relatively short time frame. It is the first phase of a major study that will ultimately produce estimates on the core damage frequency of a pressurized water reactor (PWR) during low power and shutdown conditions. Phase 2 of the study will be guided by the Phase 1 results in order to concentrate the effort on the various plant operational states, the dominant accident sequences, and pertinent data items according to their importance to core damage frequency and risk.

Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Neymotin, L.; Diamond, D.J.; Hsu, C.J.; Bozoki, G.; Kohut, P.; Fitzpatrick, R. (Brookhaven National Lab., Upton, NY (United States)); Siu, N. (Massachusetts Inst. of Tech., Cambridge, MA (United States))

1991-01-01T23:59:59.000Z

355

Preliminary results of the PWR low power and shutdown accident frequencies program: Coarse screening analysis for Surry  

Science Conference Proceedings (OSTI)

This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was performed by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES). This coarse screening analysis was performed in support of the NRC staff`s follow-up actions subsequent to the March 20, 1990 Vogtle incident with the objective of providing high-level qualitative insights within a relatively short time frame. It is the first phase of a major study that will ultimately produce estimates on the core damage frequency of a pressurized water reactor (PWR) during low power and shutdown conditions. Phase 2 of the study will be guided by the Phase 1 results in order to concentrate the effort on the various plant operational states, the dominant accident sequences, and pertinent data items according to their importance to core damage frequency and risk.

Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Neymotin, L.; Diamond, D.J.; Hsu, C.J.; Bozoki, G.; Kohut, P.; Fitzpatrick, R. [Brookhaven National Lab., Upton, NY (United States); Siu, N. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

1991-12-31T23:59:59.000Z

356

Industry approach to seismic severe accident policy implementation  

Science Conference Proceedings (OSTI)

The Nuclear Regulatory Commission (NRC) issued a severe reactor accident policy for existing plants on August 8, 1985 which describes the formal basis by which the NRC intends to resolve issues related to potential severe reactor accidents. Examination of plant-specific vulnerabilities due to seismic and other externally initiated events was considered on a later schedule and is addressed in Supplement 4 of the NRC Generic Letter No. 88-20 and a NRC guidance document, NUREG-1407, issued in June 1991. This report was prepared to provide a coherent and effective approach for seismic severe accident review which meets the intent of Generic Letter No. 88-20, Supplement 4. The recommendations in this report provide guidance on plant review types and review implementations which is consistent with the limited-scope'' intent of systematic evaluations as described in the NRC's Severe Accident Policy Statement. In addition, to assist in implementing cost-effective modifications that reduce vulnerabilities, this report also presents specific guidelines for identification and treatment of vulnerabilities that may be used as a basis for defining closure of earthquake-related severe-accident issues. This report provides procedural instructions and guidance to support resolution of earthquake-related severe accident issues. More detailed background and technical justifications for the methods are documented elsewhere, and are referenced throughout this report as appropriate.

Reed, J.W. (Benjamin (Jack R.) and Associates, Inc., Mountain View, CA (United States)); O'Hara, T.F.; Jacobson, J.P. (Yankee Atomic Electric Co., Bolton, MA (United States)); Sewell, R.T.; Cornell, C.A. (Risk Engineering, Inc., Golden, CO (United States)); Buttemer, D.R. (Pickard, Lowe and Garrick, Inc., Encinitas, CA (United States)); Schmidt, W.R.; Freed, D.A. (MPR Associates, Inc., Washington, D

1991-11-01T23:59:59.000Z

357

AIRBORNE, OPTICAL REMOTE SENSING OF METHANE AND ETHANE FOR NATURAL GAS PIPELINE LEAK DETECTION  

Science Conference Proceedings (OSTI)

Ophir Corporation was awarded a contract by the U. S. Department of Energy, National Energy Technology Laboratory under the Project Title ''Airborne, Optical Remote Sensing of Methane and Ethane for Natural Gas Pipeline Leak Detection'' on October 14, 2002. This six-month technical report summarizes the progress for each of the proposed tasks, discusses project concerns, and outlines near-term goals. Ophir has completed a data survey of two major natural gas pipeline companies on the design requirements for an airborne, optical remote sensor. The results of this survey are disclosed in this report. A substantial amount of time was spent on modeling the expected optical signal at the receiver at different absorption wavelengths, and determining the impact of noise sources such as solar background, signal shot noise, and electronic noise on methane and ethane gas detection. Based upon the signal to noise modeling and industry input, Ophir finalized the design requirements for the airborne sensor, and released the critical sensor light source design requirements to qualified vendors. Responses from the vendors indicated that the light source was not commercially available, and will require a research and development effort to produce. Three vendors have responded positively with proposed design solutions. Ophir has decided to conduct short path optical laboratory experiments to verify the existence of methane and absorption at the specified wavelength, prior to proceeding with the light source selection. Techniques to eliminate common mode noise were also evaluated during the laboratory tests. Finally, Ophir has included a summary of the potential concerns for project success and has established future goals.

Jerry Myers

2003-05-13T23:59:59.000Z

358

Comparison Between Sodium Nitrite & Sodium Hydroxide Spray Accident  

SciTech Connect

The purpose of this analysis is to compare the consequences of an 8 molar NaNO2 spray leak to the Tank Farm Final Safety Analysis Report (FSAR) evaluation of sprays of up to 19 molar (50%) NaOH. Four conditions were evaluated. These are: a spray during transfers from a one-inch pipe, a spray resulting from a truck tank Crack, a spray resulting from a truck tank rupture, and a spray in the 204-AR Waste Unloading Facility.

WILLIAMS, J.C.; HEY, B.E.

2001-11-07T23:59:59.000Z

359

Damage  

NLE Websites -- All DOE Office Websites (Extended Search)

in a nuclear reactor in a realistic time frame. XMAT's delivery of fragment ions at fission fragment energies achieves 25 dpahr. Reactor materials experience a range of dose...

360

RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident  

SciTech Connect

A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Keeping track of the damage  

NLE Websites -- All DOE Office Websites (Extended Search)

News Archives: News Archives: 2012 | 2011 | 2010 | 2009 2008 | 2007 | 2006 | 2005 2004 | 2003 | 2002 | 2001 2000 Subscribe to APS News rss feed Keeping track of the damage Scientists resolve long-standing mystery of ion-solid interactions Reprinted with kind permission from ScienceWise - Science Magazine of the Australian National University JANUARY 27, 2009 Bookmark and Share Dr. Patrick Kluth and Claudia Schnohr. Silica (silicon dioxide) is the most abundant mineral in the earth's crust and consequently is a core component in many rocks. It's quite common for such rocks to also contain natural traces of materials like uranium that undergo slow radioactive decay. This radioactivity produces energetic particles that smash through the surrounding silica creating tracks of localized damage in their wake.

362

Field Guide: Turbine Steam Path Damage  

Science Conference Proceedings (OSTI)

Steam path damage, particularly of blades, has long been recognized as a leading cause of steam turbine unavailability for large fossil fuel plants. Damage to steam path components by various mechanisms continues to result in significant economic impact domestically and internationally. Electric Power Research Institute (EPRI) Report TR-108943, Turbine Steam Path Damage: Theory and Practice, Volumes 1 and 2, was prepared to compile the most recent knowledge about turbine steam path damage: identifying th...

2011-12-12T23:59:59.000Z

363

SURFACE GEOPHYSICAL EXPLORATION DEVELOPING NONINVASIVE TOOLS TO MONITOR PAST LEAKS AROUND HANFORD TANK FARMS  

SciTech Connect

A characterization program has been developed at Hanford to image past leaks in and around the underground storage tank facilities. The program is based on electrical resistivity, a geophysical technique that maps the distribution of electrical properties of the subsurface. The method was shown to be immediately successful in open areas devoid of underground metallic infrastructure, due to the large contrast in material properties between the highly saline waste and the dry sandy host environment. The results in these areas, confirmed by a limited number of boreholes, demonstrate a tendency for the lateral extent of the underground waste plume to remain within the approximate footprint of the disposal facility. In infrastructure-rich areas, such as tank farms, the conventional application of electrical resistivity using small point-source surface electrodes initially presented a challenge for the resistivity method. The method was then adapted to directly use the buried infrastructure as electrodes for both transmission of electrical current and measurements of voltage. For example, steel-cased wells that surround the tanks were used as long electrodes, which helped to avoid much of the infrastructure problems. Overcoming the drawbacks of the long electrode method has been the focus of our work over the past seven years. The drawbacks include low vertical resolution and limited lateral coverage. The lateral coverage issue has been improved by supplementing the long electrodes with surface electrodes in areas devoid of infrastructure. The vertical resolution has been increased by developing borehole electrode arrays that can fit within the small-diameter drive casing of a direct push rig. The evolution of the program has led to some exceptional advances in the application of geophysical methods, including logistical deployment of the technology in hazardous areas, development of parallel processing resistivity inversion algorithms, and adapting the processing tools to accommodate electrodes of all shapes and locations. The program is accompanied by a full set of quality assurance procedures that cover the layout of sensors, measurement strategies, and software enhancements while insuring the integrity of stored data. The data have been shown to be useful in identifying previously unknown contaminant sources and defining the footprint of precipitation recharge barriers to retard the movement of existing contamination.

MYERS DA; RUCKER DF; LEVITT MT; CUBBAGE B; NOONAN GE; MCNEILL M; HENDERSON C

2011-06-17T23:59:59.000Z

364

PNNL Results from 2010 CALIBAN Criticality Accident Dosimeter Intercomparison Exercise  

SciTech Connect

This document reports the results of the Hanford personnel nuclear accident dosimeter (PNAD) and fixed nuclear accident dosimeter (FNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on September 20-23, 2010. Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimeter intercomparison exercise at the Commissariat a Energie Atomique (CEA) Valduc Center near Dijon, France on September 20-23, 2010. The intercomparison exercise was funded by the U.S. Department of Energy, Nuclear Criticality Safety Program, with Lawrence Livermore National Laboratory as the lead Laboratory. PNNL was one of six invited DOE Laboratory participants. The other participating Laboratories were: Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Savannah River Site (SRS), the Y-12 National Security Complex at Oak Ridge, and Sandia National Laboratory (SNL). The goals of PNNL's participation in the intercomparison exercise were to test and validate the procedures and algorithm currently used for the Hanford personnel nuclear accident dosimeters (PNADs) on the metallic reactor, CALIBAN, to test exposures to PNADs from the side and from behind a phantom, and to test PNADs that were taken from a historical batch of Hanford PNADs that had varying degrees of degradation of the bare indium foil. Similar testing of the PNADs was done on the Valduc SILENE test reactor in 2009 (Hill and Conrady, 2010). The CALIBAN results are reported here.

Hill, Robin L.; Conrady, Matthew M.

2011-10-28T23:59:59.000Z

365

Ductile damage model with void coalescence  

SciTech Connect

A general model for ductile damage in metals is presented. It includes damage induced by shear stress as well as damage caused by volumetric tension. Spallation is included as a special case. Strain induced damage is also treated. Void nucleation and growth are included and give rise to strain rate effects. Strain rate effects also arise in the model through elastic release wave propagation between damage centers. Underlying physics of the model is the nucleation, growth, and coalescence of voids in a plastically flowing solid. Implementation of the model in hydrocodes is discussed.

Tonks, D.L.

1995-03-01T23:59:59.000Z

366

Chemical damage due to drilling operations  

DOE Green Energy (OSTI)

The drilling of geothermal wells can result in near wellbore damage of both the injection wells and production wells if proper precautions are not taken. Very little specific information on the chemical causes for drilling damage that can directly be applied to the drilling of a geothermal well in a given situation is available in the literature. As part of the present work, the sparse literature references related to the chemical aspects of drilling damage are reviewed. The various sources of chemically induced drilling damages that are related to drilling operations are summarized. Various means of minimizing these chemical damages during and after the drilling of a geothermal well are suggested also.

Vetter, O.J.; Kandarpa, V.

1982-07-14T23:59:59.000Z

367

An application of probabilistic safety assessment methods to model aircraft systems and accidents  

DOE Green Energy (OSTI)

A case study modeling the thrust reverser system (TRS) in the context of the fatal accident of a Boeing 767 is presented to illustrate the application of Probabilistic Safety Assessment methods. A simplified risk model consisting of an event tree with supporting fault trees was developed to represent the progression of the accident, taking into account the interaction between the TRS and the operating crew during the accident, and the findings of the accident investigation. A feasible sequence of events leading to the fatal accident was identified. Several insights about the TRS and the accident were obtained by applying PSA methods. Changes proposed for the TRS also are discussed.

Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

1998-08-01T23:59:59.000Z

368

Accident Investigation at the Idaho National Laboratory Engineering  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Accident Investigation at the Idaho National Laboratory Engineering Accident Investigation at the Idaho National Laboratory Engineering Demonstration Facility, February 2013 Accident Investigation at the Idaho National Laboratory Engineering Demonstration Facility, February 2013 On Monday, February 12, 2013, a principal investigator at the Idaho National Laboratory (INL) Engineering Demonstration Facility (IEDF) was testing the system configuration of experimental process involving liquid sodium carbonate. An unanticipated event occurred that resulted in the ejection of the 900° C liquid sodium carbonate from the system. The ejected liquid came into contact with the principal investigator and caused multiple second and third degree burn injuries to approximately 10 percent of his body. The Office of Health, Safety and Security (HSS) Site Lead for

369

Sec. Herrington Leads Delegation in Response to Chernobyl Accident |  

National Nuclear Security Administration (NNSA)

Sec. Herrington Leads Delegation in Response to Chernobyl Accident | Sec. Herrington Leads Delegation in Response to Chernobyl Accident | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Sec. Herrington Leads Delegation in Response to ... Sec. Herrington Leads Delegation in Response to Chernobyl Accident

370

Comparison of fusedose and MACCS2 accident dose codes  

Science Conference Proceedings (OSTI)

The purpose of this paper is to present and document the differences discovered when comparing the two accident dose codes FUSEDOSE and MACCS2. Each code`s methodology is first discussed. With this background, the important comparison parameters are discussed and the resulting differences are presented. It is not the purpose of this paper to draw conclusions as to which code is more reliable but, it is hoped that the data presented will help in deciding upon further actions to be taken, if at all, to improve accident dose calculations. 7 refs., 1 fig., 1 tab.

Sevigny, L.M. [Univ. of California, Berkeley, CA (United States)

1996-12-31T23:59:59.000Z

371

Interface modeling to predict well casing damage for big hill strategic petroleum reserve.  

SciTech Connect

Oil leaks were found in well casings of Caverns 105 and 109 at the Big Hill Strategic Petroleum Reserve site. According to the field observations, two instances of casing damage occurred at the depth of the interface between the caprock and top of salt. This damage could be caused by interface movement induced by cavern volume closure due to salt creep. A three dimensional finite element model, which allows each cavern to be configured individually, was constructed to investigate shear and vertical displacements across each interface. The model contains interfaces between each lithology and a shear zone to examine the interface behavior in a realistic manner. This analysis results indicate that the casings of Caverns 105 and 109 failed by shear stress that exceeded shear strength due to the horizontal movement of the top of salt relative to the caprock, and tensile stress due to the downward movement of the top of salt from the caprock, respectively. The casings of Caverns 101, 110, 111 and 114, located at the far ends of the field, are predicted to be failed by shear stress in the near future. The casings of inmost Caverns 107 and 108 are predicted to be failed by tensile stress in the near future.

Ehgartner, Brian L.; Park, Byoung Yoon

2012-02-01T23:59:59.000Z

372

Assessment of extent and degree of thermal damage to polymeric materials in the Three Mile Island Unit 2 Reactor building  

DOE Green Energy (OSTI)

This paper describes assumptions and procedures used to perform thermal damage analysis caused by post loss-of-coolant-accident (LOCA) hydrogen deflagration at Three Mile Island Unit 2 Reactor. Examination of available photographic evidence yields data on the extent and range of thermal and burn damage. Thermal damage to susceptible material in accessible regions of the reactor building was distributed in non-uniform patterns. No clear explanation for non-uniformity was found in examined evidence, e.g., burned materials were adjacent to materials that appear similar but were not burned. Because these items were in proximity to vertical openings that extend the height of the reactor building, we assume the unburned materials preferentially absorbed water vapor during periods of high, local steam concentration. A control pendant from the polar crane located in the top of the reactor building sustained asymmetric burn damage of decreasing degree from top to bottom. Evidence suggests the polar-crane pendant side that experienced heaviest damage was exposed to intense radiant energy from a transient fire plume in the reactor containment volume. Simple hydrogen-fire-exposure tests and heat transfer calculations approximate the degree of damage found on inspected materials from the containment building and support for an estimated 8% pre-fire hydrogen.

Alvares, N.J.

1985-06-01T23:59:59.000Z

373

Bayesian hierarchical models for soil CO{sub 2} flux and leak detection at geologic sequestration sites  

Science Conference Proceedings (OSTI)

Proper characterizations of background soil CO{sub 2} respiration rates are critical for interpreting CO{sub 2} leakage monitoring results at geologic sequestration sites. In this paper, a method is developed for determining temperature-dependent critical values of soil CO{sub 2} flux for preliminary leak detection inference. The method is illustrated using surface CO{sub 2} flux measurements obtained from the AmeriFlux network fit with alternative models for the soil CO{sub 2} flux versus soil temperature relationship. The models are fit first to determine pooled parameter estimates across the sites, then using a Bayesian hierarchical method to obtain both global and site-specific parameter estimates. Model comparisons are made using the deviance information criterion (DIC), which considers both goodness of fit and model complexity. The hierarchical models consistently outperform the corresponding pooled models, demonstrating the need for site-specific data and estimates when determining relationships for background soil respiration. A hierarchical model that relates the square root of the CO{sub 2} flux to a quadratic function of soil temperature is found to provide the best fit for the AmeriFlux sites among the models tested. This model also yields effective prediction intervals, consistent with the upper envelope of the flux data across the modeled sites and temperature ranges. Calculation of upper prediction intervals using the proposed method can provide a basis for setting critical values in CO{sub 2} leak detection monitoring at sequestration sites.

Yang, Ya-Mei; Small, Mitchell J.; Junker, Brian; Bromhal, Grant S.; Strazisar, Brian; Wells, Arthur

2011-10-01T23:59:59.000Z

374

A SCOPING STUDY: Development of Probabilistic Risk Assessment Models for Reactivity Insertion Accidents During Shutdown In U.S. Commercial Light Water Reactors  

SciTech Connect

This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of a small set of comprehensive event trees and fault trees and recommendation for future work.

S. Khericha

2011-06-01T23:59:59.000Z

375

Substantiation of Thermodynamic Criteria of Explosion Safety in Process of Severe Accidents in Pressure Vessel Reactors  

E-Print Network (OSTI)

The paper represents original development of thermodynamic criteria of occurrence conditions of steam-gas explosions in the process of severe accidents. The received results can be used for modelling of processes of severe accidents in pressure vessel reactors.

Skalozubov, V I; Jarovoj, S S; Kochnyeva, V Yu

2012-01-01T23:59:59.000Z

376

Substantiation of Thermodynamic Criteria of Explosion Safety in Process of Severe Accidents in Pressure Vessel Reactors  

E-Print Network (OSTI)

The paper represents original development of thermodynamic criteria of occurrence conditions of steam-gas explosions in the process of severe accidents. The received results can be used for modelling of processes of severe accidents in pressure vessel reactors.

V. I. Skalozubov; V. N. Vashchenko; S. S. Jarovoj; V. Yu. Kochnyeva

2012-03-27T23:59:59.000Z

377

Three dimensional effects in analysis of PWR steam line break accident  

E-Print Network (OSTI)

A steam line break accident is one of the possible severe abnormal transients in a pressurized water reactor. It is required to present an analysis of a steam line break accident in the Final Safety Analysis Report (FSAR) ...

Tsai, Chon-Kwo

378

Report on the Scope of the Accident Investigation of the Tristan...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Report on the Scope of the Accident Investigation of the Tristan Fire at the DOE Brookhaven National Laboratory, IG-0386 Report on the Scope of the Accident Investigation of the...

379

Combining neural methods and knowledge-based methods in accident management  

Science Conference Proceedings (OSTI)

Accident management became a popular research issue in the early 1990s. Computerized decision support was studied from many points of view. Early fault detection and information visualization are important key issues in accident management also today. ...

Miki Sirola, Jaakko Talonen

2012-01-01T23:59:59.000Z

380

Nanofoams Response to Radiation Damage  

Science Conference Proceedings (OSTI)

Conclusions of this presentation are: (1) np-Au foams were successfully synthesized by de-alloying process; (2) np-Au foams remain porous structure after Ne ion irradiation to 1 dpa; (3) SFTs were observed in irradiated np-Au foams with highest and intermediate flux, while no SFTs were observed with lowest flux; (4) SFTs were observed in irradiated np-Au foams at RT, whereas no SFTs were observed at LNT irradiation; (5) The diffusivity of vacancies in Au at RT is high enough so that the vacancies have enough time to agglomerate and thus collapse. As a result, SFTs were formed; (6) The high flux created much more damage/time, vacancies don't have enough time to diffuse or recombine. As a result, SFTs were formed.

Fu, Engang [Los Alamos National Laboratory; Serrano De Caro, Magdalena [Los Alamos National Laboratory; Wang, Yongqiang [Los Alamos National Laboratory; Nastasi, Michael [Nebraska Center for Energy Sciences Research, University of Nebraska-Lincoln, NE 68508; Zepeda-Ruiz, Luis [PLS, Lawrence Livermore National Laboratory, Livermore, CA 94551; Bringa, Eduardo M. [CONICET and Inst. Ciencias Basicas, Universidad Nacional de Cuyo, Mendoza, 5500 Argentina; Baldwin, Jon K. [Los Alamos National Laboratory; Caro, Jose A. [Los Alamos National Laboratory

2012-07-30T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

THE ANALYSIS OF FATAL ACCIDENTS IN INDIAN D. Sengupta1  

E-Print Network (OSTI)

THE ANALYSIS OF FATAL ACCIDENTS IN INDIAN COAL MINES A. Mandal D. Sengupta1 Indian Statistical of Indian coal mines from April 1989 to March 1998. It is found that Indian mines have considerably higher over 600,000 miners and other workers. Safety in the Indian coal mines is therefore a very important

Mandal, Abhyuday

382

A SUMMARY OF INDUSTRIAL ACCIDENTS IN USAEC FACILITIES  

SciTech Connect

The accident experience of the AEC contractor operation for 1959 and 1960 is reported. Incidents involving radio active materials are described. A table of inadvertent criticality was included to supplement other tables. A tabulation of exposure records at values from 0 to 15 r is given. (M.C.G.)

1962-10-31T23:59:59.000Z

383

Normal accidents: Data quality problems in ERP-enabled manufacturing  

Science Conference Proceedings (OSTI)

The efficient operation of Enterprise Resource Planning (ERP) systems largely depends on data quality. ERP can improve data quality and information sharing within an organization. It can also pose challenges to data quality. While it is well known that ... Keywords: Data quality, ERP, complexity, enterprise resource planning, normal accident, tight coupling

Lan Cao, Hongwei Zhu

2013-05-01T23:59:59.000Z

384

Hanford Waste Tank Bump Accident and Consequence Analysis  

Science Conference Proceedings (OSTI)

This report provides a new evaluation of the Hanford tank bump accident analysis and consequences for incorporation into the Authorization Basis. The analysis scope is for the safe storage of waste in its current configuration in single-shell and double-shell tanks.

BRATZEL, D.R.

2000-06-20T23:59:59.000Z

385

Criticality accident alarm system at the Fernald Environmental Management Project  

SciTech Connect

The purpose of this paper is to give a description of the Criticality Accident Alarm System (CAAS) presently installed at the Fernald Environmental Management Project (FEMP) for monitoring areas requiring criticality controls, and some of the concerns associated with the operation of this system. The system at the FEMP is known as the Radiation Detection Alarm (RDA) System.

Marble, R.C.; Brown, T.D.; Wooldridge, J.C.

1994-06-01T23:59:59.000Z

386

Test Data for USEPR Severe Accident Code Validation  

SciTech Connect

This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: Fuel Heatup and Melt Progression Reactor Coolant System (RCS) Thermal Hydraulics In-Vessel Molten Pool Formation and Heat Transfer Fuel/Coolant Interactions during Relocation Debris Heat Loads to the Vessel Vessel Failure Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure Melt Spreading and Coolability Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

J. L. Rempe

2007-05-01T23:59:59.000Z

387

Getting to necessary and sufficient-developing accident scenarios for risk assessment  

Science Conference Proceedings (OSTI)

This paper presents a simple, systematic approach for developing accident scenarios using generic accident types. Result is a necessary and sufficient set of accident scenarios that can be used to establish the safety envelope for a facility or operation. Us of this approach along with the methodology of SAND95-0320 will yield more consistent accident analyses between facilities and provide a sound basis for allocating limited risk reduction resources.

Mahn, J.A.

1996-05-01T23:59:59.000Z

388

Radionuclide release calculations for selected severe accident scenarios  

Science Conference Proceedings (OSTI)

This report provides the results of source term calculations that were performed in support of the NUREG-1150 study. Severe Accident Risks: An Assessment for Five US Nuclear Power Plants.'' This is the sixth volume of a series of reports. It supplements results presented in the earlier volumes. Analyses were performed for three of the NUREG-1150 plants: Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Sequoyah, an ice condenser containment, pressurized water reactor. Complete source term results are presented for the following sequences: short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA for the Surry plant; station blackout with a pump seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant. In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCH3 analyses were performed for the Surry and Sequoyah plants to evaluate the effects of alternative emergency operating procedures involving primary and secondary depressurization on the progress of the accident. Only thermal-hydraulic results are provided for these analyses. In addition, three accident sequences were analyzed for the Surry plant for accident-induced failure of steam generator tubes. In these analyses, only the transport of radionuclides within the primary system and failed steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough. 17 refs., 176 figs., 113 tabs.

Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A. (Battelle Columbus Div., OH (USA))

1990-08-01T23:59:59.000Z

389

A SUMMARY OF INDUSTRIAL ACCIDENTS IN USAEC FACILITIES, 1961-1962  

SciTech Connect

Information is presented on accidents andd incidents occurring during 1961 and 1962 in plants owned and operated by the AEC. Revised reporting requirements established by the AEC in April 1962 are outlined. Data are summarized on radiation exposure of AEC contractor personnel, accidents involving radioactive materials, andd accidents involving fatalities. (C.H.)

1964-10-31T23:59:59.000Z

390

Detailed Analysis of In-Vessel Melt Progression in the Loss of Coolant Accident of OPR1000  

Science Conference Proceedings (OSTI)

An in-vessel severe accident progression has been analyzed to generate the basic data for an evaluation of the in-vessel severe accident management strategies and to identify the thermal hydraulic condition of the reactor vessel and the damage state of the in-vessel materials at a reactor vessel failure by using the SCDAP/RELAP5/MOD3.3 computer code during the Loss Of Coolant Accident (LOCA) without the Safety Injection (SI) of the OPR (Optimized Pressurize Reactor) 1000. Best estimate calculation of the small break LOCAs of 1.35 inch and 2 inch, the medium break LOCAs of 3 inch and a 4.28 inch, and a large break LOCA of 9.8 inch without the SI have been performed from a transient initiation to a reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that in all the transients, approximately 30-40 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of a reactor vessel failure. In the small and large break LOCAs, the reactor vessel failed at an early time of approximately 70-110 minutes after the transients were initiated. Since the Safety Injection Tanks (SITs) were actuated effectively in the medium break LOCAs, the reactor vessel failed at a later time of approximately 200-400 minutes after the transients were initiated. At the time of a reactor vessel failure, approximately 45-55 % of the fuel rod cladding was oxidized in the small and medium break LOCAs. However, approximately 20 % of the fuel rod cladding was oxidized because of a coolant loss through the break in the large break LOCA of the OPR1000. (authors)

Park, R.J.; Kim, S.B.; Kim, H.D. [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2006-07-01T23:59:59.000Z

391

WEB RESOURCE: Radiation Damage in Materials - TMS  

Science Conference Proceedings (OSTI)

Feb 12, 2007 ... Topic Title: WEB RESOURCE: Radiation Damage in Materials Topic Summary: Allen, Todd. Lecture notes for Univ. of Wisconsin course

392

Detection and Location of Damage on Pipelines  

SciTech Connect

The INEEL has developed and successfully tested a real-time pipeline damage detection and location system. This system uses porous metal resistive traces applied to the pipe to detect and locate damage. The porous metal resistive traces are sprayed along the length of a pipeline. The unique nature and arrangement of the traces allows locating the damage in real time along miles of pipe. This system allows pipeline operators to detect damage when and where it is occurring, and the decision to shut down a transmission pipeline can be made with actual real-time data, instead of conservative estimates from visual inspection above the area.

Karen A. Moore; Robert Carrington; John Richardson

2003-11-01T23:59:59.000Z

393

Schools - Electronic Equipment Damage Due to Lightning  

Science Conference Proceedings (OSTI)

This power quality (PQ) case study presents the investigation of damage to a school's phone equipment, security alarm, and network computer system during a lightning storm.

2003-12-31T23:59:59.000Z

394

Wind Damage in Washington, DC, 1975  

Science Conference Proceedings (OSTI)

... a large region in the Eastern United States was subjected to severe winds. ... Bureau of Standards (NBS) conducted a limited study of wind damage to ...

2011-08-12T23:59:59.000Z

395

Natural Resource Damages Assessment | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

Council Report Calls for Official Natural Resource Damage Assessment for LANL NRDA PreAssessment Screen, January 2010 Factsheet: Los Alamos National Laboratory...

396

NDE, Foreign Object Damage, and Modeling  

Science Conference Proceedings (OSTI)

Oct 10, 2012 ... Ceramic Matrix Composites: NDE, Foreign Object Damage, and .... used in cylindrical liners, pistons, rings and combustion chamber for...

397

Analysis of Underground Storage Tanks System Materials to Increased Leak Potential Associated with E15 Fuel  

DOE Green Energy (OSTI)

The Energy Independence and Security Act (EISA) of 2007 was enacted by Congress to move the nation toward increased energy independence by increasing the production of renewable fuels to meet its transportation energy needs. The law establishes a new renewable fuel standard (RFS) that requires the nation to use 36 billion gallons annually (2.3 million barrels per day) of renewable fuel in its vehicles by 2022. Ethanol is the most widely used renewable fuel in the US, and its production has grown dramatically over the past decade. According to EISA and RFS, ethanol (produced from corn as well as cellulosic feedstocks) will make up the vast majority of the new renewable fuel requirements. However, ethanol use limited to E10 and E85 (in the case of flex fuel vehicles or FFVs) will not meet this target. Even if all of the E0 gasoline dispensers in the country were converted to E10, such sales would represent only about 15 billion gallons per year. If 15% ethanol, rather than 10% were used, the potential would be up to 22 billion gallons. The vast majority of ethanol used in the United States is blended with gasoline to create E10, that is, gasoline with up to 10% ethanol. The remaining ethanol is sold in the form of E85, a gasoline blend with as much as 85% ethanol that can only be used in FFVs. Although DOE remains committed to expanding the E85 infrastructure, that market will not be able to absorb projected volumes of ethanol in the near term. Given this reality, DOE and others have begun assessing the viability of using intermediate ethanol blends as one way to transition to higher volumes of ethanol. In October of 2010, the EPA granted a partial waiver to the Clean Air Act allowing the use of fuel that contains up to 15% ethanol for the model year 2007 and newer light-duty motor vehicles. This waiver represents the first of a number of actions that are needed to move toward the commercialization of E15 gasoline blends. On January 2011, this waiver was expanded to include model year 2001 light-duty vehicles, but specifically prohibited use in motorcycles and off-road vehicles and equipment. UST stakeholders generally consider fueling infrastructure materials designed for use with E0 to be adequate for use with E10, and there are no known instances of major leaks or failures directly attributable to ethanol use. It is conceivable that many compatibility issues, including accelerated corrosion, do arise and are corrected onsite and, therefore do not lead to a release. However, there is some concern that higher ethanol concentrations, such as E15 or E20, may be incompatible with current materials used in standard gasoline fueling hardware. In the summer of 2008, DOE recognized the need to assess the impact of intermediate blends of ethanol on the fueling infrastructure, specifically located at the fueling station. This includes the dispenser and hanging hardware, the underground storage tank, and associated piping. The DOE program has been co-led and funded by the Office of the Biomass Program and Vehicle Technologies Program with technical expertise from the Oak Ridge National Laboratory (ORNL) and the National Renewable Energy Laboratory (NREL). The infrastructure material compatibility work has been supported through strong collaborations and testing at Underwriters Laboratories (UL). ORNL performed a compatibility study investigating the compatibility of fuel infrastructure materials to gasoline containing intermediate levels of ethanol. These results can be found in the ORNL report entitled Intermediate Ethanol Blends Infrastructure Materials Compatibility Study: Elastomers, Metals and Sealants (hereafter referred to as the ORNL intermediate blends material compatibility study). These materials included elastomers, plastics, metals and sealants typically found in fuel dispenser infrastructure. The test fuels evaluated in the ORNL study were SAE standard test fuel formulations used to assess material-fuel compatibility within a relatively short timeframe. Initially, these material studies included test fuels of Fuel C,

Kass, Michael D [ORNL; Theiss, Timothy J [ORNL; Janke, Christopher James [ORNL; Pawel, Steven J [ORNL

2012-07-01T23:59:59.000Z

398

Analysis of Underground Storage Tanks System Materials to Increased Leak Potential Associated with E15 Fuel  

Science Conference Proceedings (OSTI)

The Energy Independence and Security Act (EISA) of 2007 was enacted by Congress to move the nation toward increased energy independence by increasing the production of renewable fuels to meet its transportation energy needs. The law establishes a new renewable fuel standard (RFS) that requires the nation to use 36 billion gallons annually (2.3 million barrels per day) of renewable fuel in its vehicles by 2022. Ethanol is the most widely used renewable fuel in the US, and its production has grown dramatically over the past decade. According to EISA and RFS, ethanol (produced from corn as well as cellulosic feedstocks) will make up the vast majority of the new renewable fuel requirements. However, ethanol use limited to E10 and E85 (in the case of flex fuel vehicles or FFVs) will not meet this target. Even if all of the E0 gasoline dispensers in the country were converted to E10, such sales would represent only about 15 billion gallons per year. If 15% ethanol, rather than 10% were used, the potential would be up to 22 billion gallons. The vast majority of ethanol used in the United States is blended with gasoline to create E10, that is, gasoline with up to 10% ethanol. The remaining ethanol is sold in the form of E85, a gasoline blend with as much as 85% ethanol that can only be used in FFVs. Although DOE remains committed to expanding the E85 infrastructure, that market will not be able to absorb projected volumes of ethanol in the near term. Given this reality, DOE and others have begun assessing the viability of using intermediate ethanol blends as one way to transition to higher volumes of ethanol. In October of 2010, the EPA granted a partial waiver to the Clean Air Act allowing the use of fuel that contains up to 15% ethanol for the model year 2007 and newer light-duty motor vehicles. This waiver represents the first of a number of actions that are needed to move toward the commercialization of E15 gasoline blends. On January 2011, this waiver was expanded to include model year 2001 light-duty vehicles, but specifically prohibited use in motorcycles and off-road vehicles and equipment. UST stakeholders generally consider fueling infrastructure materials designed for use with E0 to be adequate for use with E10, and there are no known instances of major leaks or failures directly attributable to ethanol use. It is conceivable that many compatibility issues, including accelerated corrosion, do arise and are corrected onsite and, therefore do not lead to a release. However, there is some concern that higher ethanol concentrations, such as E15 or E20, may be incompatible with current materials used in standard gasoline fueling hardware. In the summer of 2008, DOE recognized the need to assess the impact of intermediate blends of ethanol on the fueling infrastructure, specifically located at the fueling station. This includes the dispenser and hanging hardware, the underground storage tank, and associated piping. The DOE program has been co-led and funded by the Office of the Biomass Program and Vehicle Technologies Program with technical expertise from the Oak Ridge National Laboratory (ORNL) and the National Renewable Energy Laboratory (NREL). The infrastructure material compatibility work has been supported through strong collaborations and testing at Underwriters Laboratories (UL). ORNL performed a compatibility study investigating the compatibility of fuel infrastructure materials to gasoline containing intermediate levels of ethanol. These results can be found in the ORNL report entitled Intermediate Ethanol Blends Infrastructure Materials Compatibility Study: Elastomers, Metals and Sealants (hereafter referred to as the ORNL intermediate blends material compatibility study). These materials included elastomers, plastics, metals and sealants typically found in fuel dispenser infrastructure. The test fuels evaluated in the ORNL study were SAE standard test fuel formulations used to assess material-fuel compatibility within a relatively short timeframe. Initially, these material studies included test fuels of Fuel C,

Kass, Michael D [ORNL; Theiss, Timothy J [ORNL; Janke, Christopher James [ORNL; Pawel, Steven J [ORNL

2012-07-01T23:59:59.000Z

399

Double-Shell Tank Visual Inspection Changes REsulting from the Tank 241-AY-102 Primary Tank Leak - 14193  

SciTech Connect

As part of the Double-Shell Tank (DST) Integrity Program, remote visual inspections are utilized to perform qualitative in-service inspections of the DSTs in order to provide a general overview of the condition of the tanks. During routine visual inspections of tank 241-AY -1 02 (A Y -1 02) in August 2012, anomalies were identified on the annulus floor which resulted in further evaluations. In October 2012, Washington River Protection Solutions, LLC determined that the primary tank of AY -102 was leaking. Following identification of the tank AY-102 probable leak cause, evaluations considered the adequacy of the existing annulus inspection frequency with respect to the circumstances of the tank AY-1021eak and the advancing age of the DST structures. The evaluations concluded that the interval between annulus inspections should be shortened for all DSTs, and each annulus inspection should cover > 95 percent of annulus floor area, and the portion of the primary tank (i.e., dome, sidewall, lower knuckle, and insulating refractory) that is visible from the annulus inspection risers. In March 2013, enhanced visual inspections were performed for the six oldest tanks: 241-AY-101, 241-AZ-101,241-AZ-102, 241-SY-101, 241-SY-102, and 241-SY-103, and no evidence of leakage from the primary tank were observed. Prior to October 2012, the approach for conducting visual examinations of DSTs was to perform a video examination of each tank's interior and annulus regions approximately every five years (not to exceed seven years between inspections). Also, the annulus inspection only covered about 42 percent of the annulus floor.

Girardot, Crystal L.; Washenfelder, Dennis J.; Johnson, Jeremy M.; Engeman, Jason K.

2013-11-14T23:59:59.000Z

400

Damage from methamphetamine abuse documented  

NLE Websites -- All DOE Office Websites (Extended Search)

Dennis Tartaglia, 212 481-7000, dennist@mbooth.com or Karen McNulty Walsh, 631 344-8350 go to home page Dennis Tartaglia, 212 481-7000, dennist@mbooth.com or Karen McNulty Walsh, 631 344-8350 go to home page 01-16 March 1, 2001 Researchers Document Brain Damage, Reduction in Motor and Cognitive Function from Methamphetamine Abuse "Speed" Shows More Neurotoxic Effects Than Heroin, Cocaine, or Alcohol UPTON, NY -- Two studies by researchers at the U.S. Department of Energy's Brookhaven National Laboratory provide evidence for the first time that abuse of methamphetamine -- the drug commonly known as "speed" -- is associated with physiological changes in two systems of the human brain. The changes are evident even for abusers who have not taken the drug for a year or more. The studies also found that methamphetamine abusers have reduced cognitive and motor functions, even at one year after quitting the drug. The findings appear in the March issue of the American Journal of Psychiatry.

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors  

Science Conference Proceedings (OSTI)

The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).

Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

1987-01-01T23:59:59.000Z

402

Mitigation of Severe Accident Consequences Using Inherent Safety Principles  

Science Conference Proceedings (OSTI)

Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to automatically respond to the change in reactor conditions and to result in a benign response to these events. This approach has the advantage of being relatively simple to implement, and does not face the issue of reliability since only fundamental physical phenomena are used in a passive manner, not active engineered systems. However, the challenge is to present a convincing case that such passive means can be implemented and used. The purpose of this paper is to describe this third approach in detail, the technical basis and experimental validation for the approach, and the resulting reactor performance that can be achieved for ATWS events.

R. A. Wigeland; J. E. Cahalan

2009-12-01T23:59:59.000Z

403

A plastic damage approach for confined concrete  

Science Conference Proceedings (OSTI)

There are many situations in which it is necessary to increase the capacity of structures in use. This need maybe either for a change of use or because the structures have suffered some damage or have shown little resistance in case of extreme loads ... Keywords: Concrete, Confinement, Damage, Dilation, Fiber reinforced composites, Plasticity

B. M. Luccioni; V. C. Rougier

2005-10-01T23:59:59.000Z

404

Portsmouth Site Plant Surpasses Five Years Without Lost-Time Accident |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Plant Surpasses Five Years Without Lost-Time Plant Surpasses Five Years Without Lost-Time Accident Portsmouth Site Plant Surpasses Five Years Without Lost-Time Accident November 26, 2013 - 12:00pm Addthis BWCS employees from all departments of the DUF6 project at the Portsmouth site come together to mark five years without a lost-time accident. BWCS employees from all departments of the DUF6 project at the Portsmouth site come together to mark five years without a lost-time accident. Russ Hall, environment, safety and health supervisor, changes the DUF6 project sign to mark five years without a lost-time accident. Russ Hall, environment, safety and health supervisor, changes the DUF6 project sign to mark five years without a lost-time accident. BWCS employees from all departments of the DUF6 project at the Portsmouth site come together to mark five years without a lost-time accident.

405

Order Module--DOE Order 225.1B, ACCIDENT INVESTIGATIONS | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Order 225.1B, ACCIDENT INVESTIGATIONS Order 225.1B, ACCIDENT INVESTIGATIONS Order Module--DOE Order 225.1B, ACCIDENT INVESTIGATIONS DOE O 225.1B prescribes organizational responsibilities, authorities, and requirements for conducting investigations of certain accidents occurring at DOE sites, facilities, areas, operations, and activities. The purpose of the accident investigation is to understand and identify the causes that contributed to the accident so those deficiencies can be addressed and corrected. This, in turn, is intended to prevent recurrence and promote improved environmental protection and safety and health of DOE employees, contractors, and the public. Moreover, accident investigations are used to promote the values and concepts of a learning organization. The department's integrated safety management (ISM) feedback and improvement

406

Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Natural Gas Safety Natural Gas Safety after a Traffic Accident to someone by E-mail Share Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Facebook Tweet about Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Twitter Bookmark Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Google Bookmark Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Delicious Rank Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on Digg Find More places to share Alternative Fuels Data Center: Natural Gas Safety after a Traffic Accident on AddThis.com... More in this section... Natural Gas Basics Benefits & Considerations Stations Vehicles Availability Conversions Emissions

407

Volume II - Accident and Operational Safety Analysis Handbook  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

208-2012 208-2012 July 2012 DOE HANDBOOK Accident and Operational Safety Analysis Volume II: Operational Safety Analysis Techniques U.S. Department of Energy Washington, D.C. 20585 NOT MEASUREMENT SENSITIVE DOE-HDBK-1208-2012 i ACKNOWLEDGEMENTS This Department of Energy (DOE) Accident and Operational Safety Analysis Handbook was prepared under the sponsorship of the DOE Office of Health Safety and Security (HSS), Office of Corporate Safety Programs, and the Energy Facility Contractors Operating Group (EFCOG), Industrial Hygiene and Safety Sub-group of the Environmental Health and Safety Working Group. The preparers would like to gratefully acknowledge the authors whose works are referenced in this document, and the individuals who provided valuable technical insights and/or specific

408

The Nevada railroad system: Physical, operational, and accident characteristics  

Science Conference Proceedings (OSTI)

This report provides a description of the operational and physical characteristics of the Nevada railroad system. To understand the dynamics of the rail system, one must consider the system`s physical characteristics, routing, uses, interactions with other systems, and unique operational characteristics, if any. This report is presented in two parts. The first part is a narrative description of all mainlines and major branchlines of the Nevada railroad system. Each Nevada rail route is described, including the route`s physical characteristics, traffic type and volume, track conditions, and history. The second part of this study provides a more detailed analysis of Nevada railroad accident characteristics than was presented in the Preliminary Nevada Transportation Accident Characterization Study (DOE, 1990).

NONE

1991-09-01T23:59:59.000Z

409

Review of ARAC's involvement in the Titan II missile accident  

SciTech Connect

The Atmospheric Release Advisory Capability (ARAC) response to the Titan II accident near Damascus, Arkansas on 19 September 1980 entailed 12 personnel for periods ranging from 2 to 12 hours. The first call was a NEST Standby alert at 0415L (PCT), followed by a request for dispersal calculations at 0615L, personnel callout at 0630L, crude estimates of plausible source term scenarios at 0845-0900L, first model calculations at 1130L and final model calculations at 1500L. While several new firsts were recorded for ARAC, demonstrating expanded capabilities for NEST-type responses, time lines were very long, essential information was very scant to non-existent, and useful communication of final calculations to the accident site impossible. A detailed chronology is found in Appendix A and a list of acronyms and abbreviations is contained in Appendix B.

Sullivan, T.J.

1980-10-01T23:59:59.000Z

410

Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report  

SciTech Connect

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: Give priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools. Give special technical emphasis and funding priorityto activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors. Report to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and Commercialization. The activities performed during the feasibility assessment phase include laboratory scale experiments; fuel performance code updates; and analytical assessment of economic, operational, safety, fuel cycle, and environmental impacts of the new concepts. The development and qualification stage will consist of fuel fabrication and large scale irradiation and safety basis testing, leading to qualification and ultimate NRC licensing of the new fuel. The commercialization phase initiates technology transfer to industry for implementation. Attributes for fuels with enhanced accident tolerance include improved reaction kinetics with steam and slower hydrogen generation rate, while maintaining acceptable cladding thermo-mechanical properties; fuel thermo-mechanical properties; fuel-clad interactions; and fission-product behavior. These attributes provide a qualitative guidance for parameters that must be considered in the development of fuels and cladding with enhanced accident tolerance. However, quantitative metrics must be developed for these attributes. To initiate the quantitative metrics development, a Light Water Reactor Enhanced Accident Tolerant Fuels Metrics Development Workshop was held October 10-11, 2012, in Germantown, Maryland. This document summarizes the structure and outcome of the two-day workshop. Questions regarding the content can be directed to Lori Braase, 208-526-7763, lori.braase@inl.gov.

Lori Braase

2013-01-01T23:59:59.000Z

411

Evironmental health policy in ukraine after the Chernobyl accident  

SciTech Connect

The 1986 accident at the Chernobyl nuclear power plant in Ukraine produced severe environmental health problems. This paper reports on the environmental health conditions in Ukraine after the accident and the health policy approaches employed to respond to the environmental conditions and health problems. Crisis conditions and a period of rapid change in Ukraine contributed to the difficulties of developing and implementing policy to address serious environmental health problems. Despite these difficulties, Ukraine is taking effective action. The paper describes the primary environmental health problem areas and the efforts taken to solve them. The effect of intense public fear of radiation on policymaking is described. The paper discusses the ability of public fear to distort health policy towards certain problems, leaving problems of greater importance with fewer resources. 35 refs., 1 fig.

Page, G.W.; Bobyleva, O.A.; Naboka, M.V. [and others

1995-09-01T23:59:59.000Z

412

Experimental results from containment piping bellows subjected to severe accident conditions. Volume 1, Results from bellows tested in `like-new` conditions  

Science Conference Proceedings (OSTI)

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted under the sponsorship of the US Nuclear Regulatory Commission at Sandia National Laboratories. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen bellows have been tested, all in the `like-new` condition. (Additional tests are planned of bellows that have been subjected to corrosion.) The tests showed that bellows are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage. The test data is presented and discussed.

Lambert, L.D.; Parks, M.B. [Sandia National Labs., Albuquerque, NM (United States)

1994-09-01T23:59:59.000Z

413

A new damage testing system for detailed evaluation of damage behavior of bulk KDP and DKDP  

SciTech Connect

We describe a new damage testing approach and instrumentation that provides quantitative measurements of bulk damage performance versus fluence for several frequencies. A major advantage of this method is that it can simultaneously provide direct information on pinpoint density and size, and beam obscuration. This allows for more accurate evaluation of material performance under operational conditions. Protocols for laser conditioning to improve damage performance can also be easily and rapidly evaluated.This damage testing approach has enabled us to perform complex experiments toward probing the fundamental mechanisms of damage initiation and conditioning.

DeMange, P; Negres, R A; Carr, C W; Radousky, H B; Demos, S G

2004-11-17T23:59:59.000Z

414

AQUATIC ASSESSMENT OF THE CHERNOBYL NUCLEAR ACCIDENT AND ITS REMEDIATION  

Science Conference Proceedings (OSTI)

This modeling study evaluated aquatic environment affected by the Chernobyl nuclear accident and the effectiveness of remediation efforts. Study results indicate that radionuclide concentrations in the Pripyat and Dnieper rivers were well above the drinking water limits immediately after the Chernobyl accident, but have decreased significantly in subsequent years due to flashing, burying, and decay. Because high concentrations of 90Sr and 137Cs, the major radionuclides affecting human health through aquatic pathways, are associated with flooding, an earthen dike was constructed along the Pripyat River in its most contaminated floodplain. The dike was successful in reducing the 90Sr influx to the river by half. A 100-m-high movable dome called the New Safe Confinement is planned to cover the Chernobyl Shelter (formally called the sarcophagus) that was erected shortly after the accident. The NSC will reduce radionuclide contamination further in these rivers and nearby groundwater; however, even if the Chernobyl Shelter collapses before the NSC is built, the resulting peak concentrations of 90Sr and 137Cs in the Dnieper River would still be below the drinking water limits.

Onishi, Yasuo; Kivva, Sergey L.; Zheleznyak, Mark J.; Voitsekhovitch, Oleg V.

2007-11-01T23:59:59.000Z

415

Cold Vacuum Drying facility design basis accident analysis documentation  

SciTech Connect

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

CROWE, R.D.

2000-08-08T23:59:59.000Z

416

ANS-8. 23: Criticality accident emergency planning and response  

SciTech Connect

A study group has been formed under the auspices of ANS-8 to examine the need for a standard on nuclear criticality accident emergency planning and response. This standard would be ANS-8.23. ANSI/ANS-8.19-1984, Administrative Practices for Nuclear Criticality Safety, provides some guidance on the subject in Section 10 titled -- Planned Response to Nuclear Criticality Accidents. However, the study group has formed a consensus that Section 10 is inadequate in that technical guidance in addition to administrative guidance is needed. The group believes that a new standard which specifically addresses emergency planning and response to a perceived criticality accident is needed. Plans for underway to request the study group be designated a writing group to create a draft of such a new standard. The proposed standard will divide responsibility between management and technical staff. Generally, management will be charged with providing the necessary elements of emergency planning such as a criticality detection and alarm system, training, safe evacuation routes and assembly areas, a system for timely accountability of personnel, and an effective emergency response organization. The technical staff, on the other hand, will be made responsible for establishing specific items such as safe and clearly posted evacuation evacuation routes and dose criteria for personnel assembly areas. The key to the question of responsibilities is that management must provide the resources for the technical staff to establish the elements of an emergency response effort.

Pruvost, N.L.

1991-06-24T23:59:59.000Z

417

Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis  

Science Conference Proceedings (OSTI)

The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

Gilles Youinou; R. Sonat Sen

2013-09-01T23:59:59.000Z

418

Analysis of PWR RCS Injection Strategy During Severe Accident  

Science Conference Proceedings (OSTI)

Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

Wang, S.-J. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, K.-S. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, S.-C. [Taiwan Power Company, Taiwan (China)

2004-05-15T23:59:59.000Z

419

Health effects of the nuclear accident at Three Mile Island  

SciTech Connect

Between March 28 and April 15, 1979 the collective dose resulting from the radioactivity released to the population living within a 50-mile radius of the Three Mile Island nuclear plant was about 2000 person-rems, less than 1% of the annual natural background level. The average dose to a person living within 5 miles of the nuclear plant was less than 10% of annual background radiation. The maximum estimated radiation dose received by any one individual in the general population (excluding the nuclear plant workers) during the accident was 70 mrem. The doses received by the general population as a result of the accident were so small that there will be no detectable additional cases of cancer, developmental abnormalities, or genetic ill-health. Three Three Mile Island nuclear workers received radiation doses of about 3 to 4 rem, exceeding maximum permissible quarterly dose of 3 rem. The major health effect of the accident at Three Mile Island was that of a pronounced demoralizing effect on the general population in the Three Mile Island area, including teenagers and mothers of preschool children and the nuclear plant workers. However, this effect proved transient in all groups studied except the nuclear workers.

Fabrikant, J.I.

1980-05-01T23:59:59.000Z

420

Internally deposited fallout from the Chernobyl reactor accident  

SciTech Connect

Measurements of fallout radioactivity were made in the thyroid region, abdomen, whole body, or urine of 96 persons who were in eastern Europe at the time of the Chernobyl reactor accident or who went there shortly afterward. The most frequently encountered radionuclides were /sup 131/I, /sup 134,137/Cs, and /sup 103/Ru//sup 103/Rh. The median /sup 131/I activity in the thyroids of 42 subjects in whom radioiodine was detected and who were in Europe when the accident began was projected as 42 nCi the day the accident began. The median total body activity of /sup 134/Cs in 40 subjects in which it was detected was 1.7 nCi upon arrival in the US. For 51 subjects with detectable /sup 137/Cs burdens, the total body activity was 4.6 nCi. The risk of fatal thyroid cancer is less than 3 x 10/sup -6/ for nearly all subjects in this series. The risk of fatal cancer from /sup 134,137/Cs for subjects with cesium exposures similar to the ones observed by us, but who remained in Europe, is estimated as 1.4 x 10/sup -6/ to 4.2 x 10/sup -5/ with 95% of the risk attributable to /sup 137/Cs. 5 refs., 4 tabs.

Schlenker, R.A.; Oltman, B.G.; Lucas, H.F.

1987-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

RAPID DAMAGE ASSESSMENT FROM HIGH RESOLUTION IMAGERY  

SciTech Connect

Disaster impact modeling and analysis uses huge volumes of image data that are produced immediately following a natural or an anthropogenic disaster event. Rapid damage assessment is the key to time critical decision support in disaster management to better utilize available response resources and accelerate recovery and relief efforts. But exploiting huge volumes of high resolution image data for identifying damaged areas with robust consistency in near real time is a challenging task. In this paper, we present an automated image analysis technique to identify areas of structural damage from high resolution optical satellite data using features based on image content.

Vijayaraj, Veeraraghavan [ORNL; Bright, Eddie A [ORNL; Bhaduri, Budhendra L [ORNL

2008-01-01T23:59:59.000Z

422

Carbon Fiber Damage in Accelerator Beam  

E-Print Network (OSTI)

Carbon fibers are commonly used as moving targets in Beam Wire Scanners. Because of their thermomechanical properties they are very resistant to particle beams. Their strength deteriorates with time due to radiation damage and low-cycle thermal fatigue. In case of high intensity beams this process can accelerate and in extreme cases the fiber is damaged during a single scan. In this work a model describing the fiber temperature, thermionic emission and sublimation is discussed. Results are compared with fiber damage test performed on SPS beam in November 2008. In conclusions the limits of Wire Scanner operation on high intensity beams are drawn.

Sapinski, M; Guerrero, A; Koopman, J; Mtral, E

2009-01-01T23:59:59.000Z

423

Schools - Lightning Causing Electric Equipment Damage  

Science Conference Proceedings (OSTI)

This power quality (PQ) case study presents the investigation of damage to a school's electronic controls and integrated circuit boards within their fire alarm and clock/bell system when there are lightning storms.

2003-12-31T23:59:59.000Z

424

Dealing with Storm-Damaged Trees  

E-Print Network (OSTI)

Many homeowners need help caring for or removing damaged trees after a natural disaster. This publication explains what a certified arborist is and how to select one. It also cautions against burning debris downed by a storm.

Kirk, Melanie; Taylor, Eric; Foster, C. Darwin

2005-10-25T23:59:59.000Z

425

Dealing with Storm-Damaged Trees (Spanish)  

E-Print Network (OSTI)

Many homeowners need help caring for or removing damaged trees after a natural disaster. This publication explains what a certified arborist is and how to select one. It also cautions against burning debris downed by a storm.

Kirk, Melanie; Taylor, Eric; Foster, C. Darwin

2007-10-08T23:59:59.000Z

426

Tornado Damage Estimation Using Polarimetric Radar  

Science Conference Proceedings (OSTI)

This study investigates the use of tornadic debris signature (TDS) parameters to estimate tornado damage severity using Norman, Oklahoma (KOUN), polarimetric radar data (polarimetric version of the Weather Surveillance Radar-1988 Doppler radar). ...

David J. Bodine; Matthew R. Kumjian; Robert D. Palmer; Pamela L. Heinselman; Alexander V. Ryzhkov

2013-02-01T23:59:59.000Z

427

Laser Damage Precursors in Fused Silica  

Science Conference Proceedings (OSTI)

There is a longstanding, and largely unexplained, correlation between the laser damage susceptibility of optical components and both the surface quality of the optics, and the presence of near surface fractures in an optic. In the present work, a combination of acid leaching, acid etching, and confocal time resolved photoluminescence (CTP) microscopy has been used to study laser damage initiation at indentation sites. The combination of localized polishing and variations in indentation loads allows one to isolate and characterize the laser damage susceptibility of densified, plastically flowed and fractured fused silica. The present results suggest that: (1) laser damage initiation and growth are strongly correlated with fracture surfaces, while densified and plastically flowed material is relatively benign, and (2) fracture events result in the formation of an electronically defective rich surface layer which promotes energy transfer from the optical beam to the glass matrix.

Miller, P; Suratwala, T; Bude, J; Laurence, T A; Shen, N; Steele, W A; Feit, M; Menapace, J; Wong, L

2009-11-11T23:59:59.000Z

428

Thin Film Femtosecond Laser Damage Competition  

SciTech Connect

In order to determine the current status of thin film laser resistance within the private, academic, and government sectors, a damage competition was started at the 2008 Boulder Damage Symposium. This damage competition allows a direct comparison of the current state of the art of high laser resistance coatings since they are tested using the same damage test setup and the same protocol. In 2009 a high reflector coating was selected at a wavelength of 786 nm at normal incidence at a pulse length of 180 femtoseconds. A double blind test assured sample and submitter anonymity so only a summary of the results are presented here. In addition to the laser resistance results, details of deposition processes, coating materials and layer count, and spectral results will also be shared.

Stolz, C J; Ristau, D; Turowski, M; Blaschke, H

2009-11-14T23:59:59.000Z

429

Assessing United States hurricane damage under different environmental conditions  

E-Print Network (OSTI)

Hurricane activity between 1979 and 2011 was studied to determine damage statistics under different environmental conditions. Hurricanes cause billions of dollars of damage every year in the United States, but damage ...

Maheras, Anastasia Francis

2012-01-01T23:59:59.000Z

430

Using Landsat to Identify Thunderstorm Damage in Agricultural Regions  

Science Conference Proceedings (OSTI)

During 12 and 18 August 1999, severe thunderstorms produced damaging winds and hail that caused an estimated $50 million in damage to agriculture in west-central Illinois. Landsat-7 imagery was obtained to determine the arealextent of damage and ...

Mace L. Bentley; Thomas L. Mote; Paporn Thebpanya

2002-03-01T23:59:59.000Z

431

Controlled ion implant damage profile for etching  

DOE Patents (OSTI)

This invention pertains to a process for etching a material such as LiNbO{sub 3} by implanting ions having a plurality of different kinetic energies in an area to be etched, and then contacting the ion implanted area with an etchant. The various energies of the ions are selected to produce implant damage substantially uniformly throughout the entire depth of the zone to be etched, thus tailoring the vertical profile of the damaged zone.

Arnold, G.W. Jr.; Ashby, C.I.H.; Brannon, P.J.

1988-08-18T23:59:59.000Z

432

Reference Poster: Turbine Bearing Damage Mechanisms  

Science Conference Proceedings (OSTI)

Damage to turbine and generator bearings accounts for a significant amount of lost generation in the power industry. There are numerous known damage mechanisms affecting these bearings, and as part of EPRIs technology transfer efforts, we have developed a reference poster. This poster provides clear, concise, and visual information for a wide variety of mechanisms and is meant to supplement related EPRI projects. By providing an overview of various issues as well as information on how to ...

2012-10-04T23:59:59.000Z

433

Formation damage in underbalanced drilling operations  

E-Print Network (OSTI)

Formation damage has long been recognized as a potential source of reduced productivity and injectivity in both horizontal and vertical wells. From the moment that the pay zone is being drilled until the well is put on production, a formation is exposed to a series of fluids and operations that can reduce its productive capacity. Any process that causes a loss in the productivity of an oil-, gas-, or water-saturated formation has a damaging effect on the reservoir. These damage mechanisms predominantly fall into three major classifications: mechanical, chemical, and biological. Underbalanced drilling operations involve drilling a portion of the wellbore at fluid pressures less than that of the target formation. This technology has been used to prevent or minimize problems associated with invasive formation damage, which often greatly reduces the productivity of oil and gas reservoirs, mainly in openhole horizontal-well applications. Underbalanced drilling is not a solution for all formation-damage problems. Damage caused by poorly designed and/or executed underbalanced drilling programs can equal or exceed that which may occur with a well-designed conventional overbalanced drilling program. Four techniques are currently available to achieve underbalanced conditions while drilling. These include using lightweight drilling fluids, injecting gas down the drillpipe, injecting gas into a parasite string, and using foam. This study provides an analysis of a number of potential damage mechanisms present when drilling underbalanced. It describes each one and its influence on the productivity of a well. Additionally it presents a general description of the different techniques that can be applied to carry out successful, cost-effective UBD operations, and discusses how these techniques may be used to reduce or eliminate formation damage.

Reyes Serpa, Carlos Alberto

2003-01-01T23:59:59.000Z

434

R-damage cassette (incorporated brass sleeve)  

SciTech Connect

The R-Damage series of ten experiments is part of a long-term collaboration with RFNC/VNIIEF in pulsed power technology. These experiments use a cylindrical configuration to study spallation damage, which allows for a natural recollection of the damaged material under proper driving conditions and post-shot collection of the damaged target material for subsequent metallographic analysis. Dynamic in-situ experimental velocimetry diagnostics are also employed. LANL is responsible for the design of the experimental load and velocimetry system. VNIIEF is responsible for the design and construction of the driving explosive magnetic generator. Eight of the experiments in the planned series have been completed. Thus far, data has been collected about failure initiation of a well-characterized material (aluminum) in a cylindrical geometry, the behavior of material recollected after damage from pressures in the damage initiation regime, and the behavior of material recollected after complete failure. The final two experiments will continue the study of material recollected after complete failure. The load assembly shown is similar to that employed in the previous two experiments, with some modiflications for easier assembly.

Griego, Jeffrey Randall [Los Alamos National Laboratory

2011-01-07T23:59:59.000Z

435

DOE Order Self Study Modules - DOE O 225.1B, Accident Investigation  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

225.1B 225.1B ACCIDENT INVESTIGATIONS DOE O 225.1B Accident Investigations Familiar Level June 2011 1 June 2011 DOE ORDER O 225.1B ACCIDENT INVESTIGATIONS FAMILIAR LEVEL OBJECTIVES Given the familiar level of this module and the resources listed below, you will be able to: 1. State the purpose of implementing U.S. Department of Energy (DOE) O 225.1B. 2. Discuss the responsibilities of the heads of field elements for accident investigations. 3. Discuss the responsibilities of the appointing official in an accident investigation. 4. Discuss the responsibilities of the Accident Investigation Board Chairperson. 5. Discuss the criteria identified in appendix A of DOE O 225.1B. Note: If you think that you can complete the practice at the end of this level without

436

Clustered DNA Damage Spectrum in Primary Human Hematopoietic...  

NLE Websites -- All DOE Office Websites (Extended Search)

94805 Villejuif Cedex France Clustered DNA Damages Induced by Low Radiation Doses Irradiation of cells with low doses of X- or -rays induces clustered damages in mammalian...

437

Reactor Loose Part Damage Assessments on Steam Generator Tube Sheets.  

E-Print Network (OSTI)

??PROCTOR, WILLIAM CYRUS. Reactor Loose Part Damage Assessments on Steam Generator Tube Sheets. (Under the direction of Joseph Michael Doster). Damage from loose parts inside (more)

Proctor, William Cyrus

2010-01-01T23:59:59.000Z

438

ABSTRACT: Ion-Induced Damage Accumulation and Electron-Beam ...  

Science Conference Proceedings (OSTI)

Jun 27, 2007... Ion-Induced Damage Accumulation and Electron-Beam-Enhanced ... damage accumulation in strontium titanate from 1.0 MeV Au irradiation

439

Los Alamos National Laboratory describes storm damage to environmental...  

NLE Websites -- All DOE Office Websites (Extended Search)

Los Alamos National Laboratory describes storm damage Los Alamos National Laboratory describes storm damage to environmental monitoring stations, canyons Stations supporting Santa...

440

Descriptions of selected accidents that have occurred at nuclear reactor facilities  

SciTech Connect

This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

Bertini, H.W.

1980-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "leaks damage accidents" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

MACCS usage at Rocky Flats Plant for consequence analysis of postulated accidents  

Science Conference Proceedings (OSTI)

The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and health effects to the public from postulated plutonium releases and from postulated criticalities. These applications were conducted to support deterministic and probabilistic accident analyses for safety analyses for safety analysis reports, radiological sabotage studies, and other regulatory requests.

Foppe, T.L.; Peterson, V.L.

1993-10-01T23:59:59.000Z

442

Descriptions of selected accidents that have occurred at nuclear reactor facilities  

SciTech Connect

This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

Bertini, H.W.

1980-04-01T23:59:59.000Z

443

Assessment of ICARE/CATHARE V1 Severe Accident Code  

SciTech Connect

The ICARE/CATHARE code system has been developed