National Library of Energy BETA

Sample records for mgr lajos grof-tisza

  1. Lajos Grof-Tisza

    Broader source: Energy.gov [DOE]

    Mr. Lajos Grof-Tisza joined the Department of Energy in 2003 and currently serves as the Director of Corporate Information Systems. As the Director of Corporate Information Systems, Mr. Grof-Tisza...

  2. MGR External Events Hazards Analysis

    SciTech Connect (OSTI)

    L. Booth

    1999-11-06

    The purpose and objective of this analysis is to apply an external events Hazards Analysis (HA) to the License Application Design Selection Enhanced Design Alternative 11 [(LADS EDA II design (Reference 8.32))]. The output of the HA is called a Hazards List (HL). This analysis supersedes the external hazards portion of Rev. 00 of the PHA (Reference 8.1). The PHA for internal events will also be updated to the LADS EDA II design but under a separate analysis. Like the PHA methodology, the HA methodology provides a systematic method to identify potential hazards during the 100-year Monitored Geologic Repository (MGR) operating period updated to reflect the EDA II design. The resulting events on the HL are candidates that may have potential radiological consequences as determined during Design Basis Events (DBEs) analyses. Therefore, the HL that results from this analysis will undergo further screening and analysis based on the criteria that apply during the performance of DBE analyses.

  3. CLASSIFICATION OF THE MGR SAFEGUARDS AND SECURITY SYSTEM

    SciTech Connect (OSTI)

    J.A. Ziegler

    1999-08-31

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) safeguards and security system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998).

  4. CLASSIFICATION OF THE MGR NON-FUEL COMPONENTS DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    J.A. Ziegler

    1999-08-31

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) non-fuel components disposal container system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998).

  5. Title: Freedom of Information Request DIR DIV NAME MGR DEP AMA

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Title: Freedom of Information Request DIR DIV NAME MGR DEP AMA FMD HRM PRO AMCP AMMS ISI PIC SES DIR DIV NAME SSD AMRC AMSE EMD OOD SED OCC OCE Riehle, Dorothy (Actionee) ORP PNSO...

  6. iManage Presentation | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    iManage Presentation iManage Presentation PDF icon Tuesday_Dallas_Ballroom_D_1425_Grof-Tisza.pdf More Documents & Publications Before the House Oversight and Government Reform Subcommittee on Technology, Information Policy, Intergovernmental Relations, and Procurement Reform FY 2009 E-Government Act Report Acquisition Planning: Revised DOE Acquisition Guide Chapter 7.1

  7. list2_eligible_multifamily_buildings_10-cfr-440-22b4ii.xls | Department of

    Energy Savers [EERE]

    iManage Presentation iManage Presentation PDF icon Tuesday_Dallas_Ballroom_D_1425_Grof-Tisza.pdf More Documents & Publications Before the House Oversight and Government Reform Subcommittee on Technology, Information Policy, Intergovernmental Relations, and Procurement Reform FY 2009 E-Government Act Report Request an iPortal Account!

    NCSL/State and Tribal Government Working Group Appendix CLOSURE FOR THE SEVENTH GENERATION A REPORT FROM THE STEWARDSHIP COMMITTEE OF THE STATE AND TRIBAL

  8. galvin | The Ames Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    galvin Ames Laboratory Profile Glen Galvin Mgr Info Tech I Simulation, Modeling, & Decision Science 1620 Howe Phone Number: 515-294-6604 Email Address: galvin

  9. hoenig | The Ames Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    hoenig Ames Laboratory Profile Douglas Hoenig Mgr Facility Serv Facilities Services 158J Metals Development Phone Number: 515-294-0930 Email Address: hoenig@ameslab.gov...

  10. grootvel | The Ames Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    grootvel Ames Laboratory Profile Mark Grootveld Mgr Facility Serv Facilities Services 158 Metals Development Phone Number: 515-294-7895 Email Address: grootveld@ameslab.gov...

  11. vdahl | The Ames Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    vdahl Ames Laboratory Profile Vincent Dahl Mgr Facilities Mnt Facilities Services Maintenance Shop Phone Number: 515-294-1746 Email Address: vdahl...

  12. Emergency Operations Training Academy | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    ... Introduction Monitoring Division Mgr Training, Adv NARAC Dispersion Modeling NARAC Web Operations Overview of Consequence Management Overview of the DOENNSA Emergency ...

  13. The effects of turbine passage on C-start behavior of salmon...

    Office of Scientific and Technical Information (OSTI)

    gap runner (MGR) turbine is predicted to have lower values for several potential fish injury mechanisms, and therefore was expected to improve turbine-passage fish survival. ...

  14. The Ames Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    515-294-5663 John Clough Mgr Ames Lab Acctg 224 TASF cloughj@ameslab.gov 515-294-5623 Cassandra Dewitt Accountant I 224 TASF dewitt@ameslab.gov 515-294-4129 Kelsey Hummer...

  15. Document: NA Actionee: Dorothy Riehie

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    MGR AMRP DEP AMRP CPD AMB AMRP RCD AMB BUD AMISE AMB FIN AMISE EMD AMB HRM AMISE OOD AMIB PRO AMISE SED AMMIS 0CC AMMIS SES OCE Riehle, Dorothy (Actionee) AMIMS ISI ORP AMMS PlC...

  16. Microsoft Word - NSDBs for LA REV 003 FINAL.doc

    Office of Scientific and Technical Information (OSTI)

    QA: QA 000-30R-MGR0-00400-000-003 September 2005 Nuclear Safety Design Bases for License ... DE-AC28-01RW12101 ENG.20050929.0005 Nuclear Safety Design Bases for License ...

  17. Document: NA Actionee: Dorothy Riehie Document Date: 02/25/2010...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Request (FOI 2010-00842) DIR DIV NAME DIR DIV NAME MGR AMRC DEP AMISE AMA EMD FMD GOD HRM SED PRO 0CC AMCP OCE Riehle, Dorothy (Actionee) AMMS ORP ISI PNSO PIC RLCI SES...

  18. RESPONSE TRACKING INFORMATION BEH W.Wagner BPM M. Redmon BPM

    Office of Legacy Management (LM)

    RESPONSE TRACKING INFORMATION BEH W.Wagner BPM M. Redmon BPM . Palau BPM P. Huber BPM S. Priest BPM BCR BFC ENVIR TECH/DATA BET ENGINEERING BET GEOTE~ BET DEPUTY PROGRAM MGR.: P. Crotwell BPM PROGRAM MANAGER: R. Harbert BPM PROJECT MANAGER: COMMUNITY RELATIONS CONSTRUCTION ENGINEERING & TECHNOLOGY BET ENVIRON SAFETY & HEALTH F~~-~~~~-~-~---+-r~-~ W/A W/O K. Renfro SAIC J. Waddell SAIC S. Heptinstall SAIC DEPUTY PROGRAM MGR: T. Patlerson SAIC PROGRAM MANAGER: MGMT. SYSTEMS: SECRETARY:

  19. Monitored Geologic Repository Life Cycle Cost Estimate Assumptions Document

    SciTech Connect (OSTI)

    R. Sweeney

    2000-03-08

    The purpose of this assumptions document is to provide general scope, strategy, technical basis, schedule and cost assumptions for the Monitored Geologic Repository (MGR) life cycle cost estimate and schedule update incorporating information from the Viability Assessment (VA), License Application Design Selection (LADS), 1999 Update to the Total System Life Cycle Cost (TSLCC) estimate and from other related and updated information. This document is intended to generally follow the assumptions outlined in the previous MGR cost estimates and as further prescribed by DOE guidance.

  20. MONITORED GEOLOGIC REPOSITORY LIFE CYCLE COST ESTIMATE ASSUMPTIONS DOCUMENT

    SciTech Connect (OSTI)

    R.E. Sweeney

    2001-02-08

    The purpose of this assumptions document is to provide general scope, strategy, technical basis, schedule and cost assumptions for the Monitored Geologic Repository (MGR) life cycle cost (LCC) estimate and schedule update incorporating information from the Viability Assessment (VA) , License Application Design Selection (LADS), 1999 Update to the Total System Life Cycle Cost (TSLCC) estimate and from other related and updated information. This document is intended to generally follow the assumptions outlined in the previous MGR cost estimates and as further prescribed by DOE guidance.

  1. Audience/Panel Discussion: Sites Lesson Learned about Activity-level Work Planning and Control Using EFCOG Work Planning and Control Guideline

    Broader source: Energy.gov [DOE]

    Slide Presentation by Donna J. Governor, Deputy Dept Mgr for Planning & Integration, Lawrence Livermore National Laboratory. Lawrence Livermore National Laboratory work planning and control lessons learned and audience/panel discussion on site's lessons learned about Activity-level Work Planning and Control using EFCOG Work Planning and Control Guideline Document.

  2. Auto-Versioning Systems Image Manager

    Energy Science and Technology Software Center (OSTI)

    2013-08-01

    The av_sys_image_mgr utility provides an interface for the creation, manipulation, and analysis of system boot images for computer systems. It is primarily intended to provide a convenient method for managing the introduction of changes to boot images for long-lived production HPC systems.

  3. INDUSTRIAL/MILITARY ACTIVITY-INITIATED ACCIDENT SCREENING ANALYSIS

    SciTech Connect (OSTI)

    D.A. Kalinich

    1999-09-27

    Impacts due to nearby installations and operations were determined in the Preliminary MGDS Hazards Analysis (CRWMS M&O 1996) to be potentially applicable to the proposed repository at Yucca Mountain. This determination was conservatively based on limited knowledge of the potential activities ongoing on or off the Nevada Test Site (NTS). It is intended that the Industrial/Military Activity-Initiated Accident Screening Analysis provided herein will meet the requirements of the ''Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants'' (NRC 1987) in establishing whether this external event can be screened from further consideration or must be included as a design basis event (DBE) in the development of accident scenarios for the Monitored Geologic Repository (MGR). This analysis only considers issues related to preclosure radiological safety. Issues important to waste isolation as related to impact from nearby installations will be covered in the MGR performance assessment.

  4. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    SciTech Connect (OSTI)

    S.O. Bader

    1999-10-18

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF 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. The 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. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.

  5. S-CHA-H-00026 Rev. A PRELIMINARY SCOPING-LEVEL HAZARD ANALYSIS

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    S-CHA-H-00026 Rev. A PRELIMINARY SCOPING-LEVEL HAZARD ANALYSIS FOR THE PROCESSING OF HGTR PEBBLE FUEL AT SRS January 2015 OFFICIAL USE ONLY May be exempt from public release under the Freedom of Information Act (5 U.S.C. 552), exemption number 4 and category Commercial/Proprietary Department of Energy review required before public release. Name/Org:_C. M. Hadden, Dep Mgr H N&CSE Date: 2/4/2015 Guidance (if applicable):___n/a___ S-CHA-H-00026 Rev. A ii DISCLAIMER This document was prepared by

  6. The Ames Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Directory: Information Systems Name Title Office Email Phone Number Mark Clarridge Sys Sup Spec IV 334 TASF clarridge@ameslab.gov 515-294-6396 Karla Concannon Sys Analyst II 334 TASF kconcann@ameslab.gov 515-294-8334 Klarida Cubacub Sys Sup Spec III 334 TASF cubacub@ameslab.gov 515-294-5325 Diane Denadel Mgr Ames Lab Info Sys 334 TASF ddenadel@ameslab.gov 515-294-1061 Fang Fang 334 TASF ffang@iastate.edu 515-294-8348 Christopher Farrington Sys Sup Spec III 334 TASF farrington@ameslab.gov

  7. Federal Technical Capability Panel

    Energy Savers [EERE]

    Updated: April 2015 1 U. S. Department of Energy and National Nuclear Security Administration Federal Technical Capability Panel Organization Name Telephone Fax E-Mail FTCP CHAIR Chair (DOE/NTC) Karen L. Boardman (505) 845-6444 (505) 845-6079 kboardman@ntc.doe.gov FTCP Deputy Dave Chaney (505) 845-4300 (505) 845-4879 david.chaney@nnsa.doe.gov FTCP Technical Standards Mgr. Jeanette Yarrington (301) 903-7030 (301) 903-3445 Jeanette.Yarrington@hq.doe.gov FTCP Program Coordinator Jeannie Lozoya

  8. MESSAGE: WIA W/O CLOSING REF CLOSING REF

    Office of Legacy Management (LM)

    MESSAGE: WIA W/O CLOSING REF CLOSING REF _ CONSTRUCTION COMPL DATE J. King SAIC J. Waddell SAIC R. Wright SAIC T. Gangwer SAIC M. Khan SAIC T. Patterson SAIC R. Tucker SAIC C.Helie SAIC K. Renfro SAIC S. Heptinstall SAIC PLEASE RETURN TO PDCC FOR CORRECTIONS MGMT. SYSTEMS: PROGRAM ADMIN.: DEPUTY PROGRAM MGR: PROJECT MANAGER: () I PROGRAM MANAGER: I I ANL: AJ. Dvorak ANL A Geisler ANL G. Maraman ANL D. Dunning ANL J. Wing BNI DIRECTOR. FSRD: L Price FSRD DEP. DIRECTOR. FSRD: W.Seay FSRD SITE

  9. Actionee: Dorothy Riehie

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    S1!4T op Document: NA (HOA) Actionee: Dorothy Riehie Document Date: 07/12/2011 Due Date: NO ACTION C. 41 r 4 Author: PEARSON R Addressee: RIEHLE DC, TE , MORRIS A Title: FOIA Request for Department of Energy Documents and Records - 07/12/2011 DIR DIV NAME DIR DIV NAME MGR AMRC DEP AMSE AMA EMD FMD OOD HRM SED PRO 0CC AMCP OCE Riehie, Dorothy (Actionee) AMMS ORP ISI PNSO PIC RLCI SES Comments: THIS DOCUMENT CONTAINS OUO INFORMATION Records Schedule Information: ADM-1.28.1 Scan?: Yes Sensitive?:

  10. Attachment A -- Deliverables.xls

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    4: Unrevised SFO Paragraphs Reissued Attachment 4: Unrevised SFO Paragraphs Reissued PDF icon Unrevised SFO Paragraphs Reissued More Documents & Publications Attachment 2: Solicitation for Offers with New and Revised Green Lease Text Attachment 1: Green Lease Policies and Procedures for Lease Acquisition 1

    B - J Deliverables Attachment A TOC Deliverables DE-AC27-08RV14800 SEC. Contract Section Description Action Timing TFP CO ESQ OPA IR/HR ORP MGR DCAA B B.2 Modify contract to obligate

  11. Attachment A -- Deliverables.xls

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    I Deliverables Attachment A TOC Deliverables DE-AC27-08RV14800 SEC. FAR/DEAR Clause Reference Description Action Timing TFP CO ESQ OPA IR/HR ORP MGR DCAA I.2 FAR 52.202-1 Definitions (JUL 2004) as supplemented by DEAR 952.202-1 (Mar 2002) Verify compliance As Required L I.3 FAR 52.203-3 Gratuities (APR 1984) Verify compliance As Required L I.4 FAR 52.203-5 Covenant Against Contingent Fees (APR 1984) Verify compliance As Required L I.5 FAR 52.203-6 Restrictions on Subcontractor Sales to the

  12. ESF Mine Power Center Platforms

    SciTech Connect (OSTI)

    T.A. Misiak

    2000-02-10

    The purpose and objective of this analysis is to structurally evaluate the existing Exploratory Studies Facility (ESF) mine power center (MPC) support frames and to design service platforms that will attach to the MPC support frames. This analysis follows the Development Plan titled ''Produce Additional Design for Title 111 Evaluation Report'' (CRWMS M&O 1999a). This analysis satisfies design recommended in the ''Title III Evaluation Report for the Surface and Subsurface Power System'' (CRWMS M&O 1999b, Section 7.6) and concurred with in the ''System Safety Evaluation of Title 111 Evaluation Reports Recommended Work'' (Gwyn 1999, Section 10.1.1). This analysis does not constitute a level-3 deliverable, a level-4 milestone, or a supporting work product. This document is not being prepared in support of the Monitored Geologic Repository (MGR) Site Recommendation (SR), Environmental Impact Statement (EIS), or License Application (LA) and should not be cited as a reference in the MGR SR, EIS, or LA.

  13. Intact and Degraded Criticality Calculations for the Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package

    SciTech Connect (OSTI)

    L.M. Montierth

    2000-09-15

    The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the U.S. Department of Energy's (DOE) Shippingport Light Water Breeder Reactor (SP LWBR) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP), which is to be placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (K{sub eff}) for intact- and degraded-mode internal configurations of the codisposal WP containing Shippingport LWBR seed-type assemblies. The results of this calculation will be used to evaluate criticality issues and support the analysis that is planed to be performed to demonstrate the viability of the codisposal concept for the MGR. This calculation is associated with the waste package design and was performed in accordance with the DOE SNF Analysis Plan for FY 2000 (See Ref. 22). The document has been prepared in accordance with the Administrative Procedure AP-3.12Q, Calculations (Ref. 23).

  14. RADIOLOGICAL RELEASES DUE TO AIR AND SILICA DUST ACTIVATION IN EMPLACEMENT DRIFTS

    SciTech Connect (OSTI)

    J.S. Tang

    2003-05-07

    The purpose of this calculation is to determine the quantity and significance of annual Monitored Geologic Repository (MGR) subsurface normal radiological releases due to neutron activation of air and silica dust in emplacement drifts. This calculation includes the following items: (1) Calculate activation of ventilation airflow through emplacement drifts to quantify radioactive gaseous releases; and (2) Calculate the bounding potential activated silica dust concentration and releases. The sources of silica dust may arise from air supply to emplacement drifts as well as host rock around emplacement drifts. For this calculation, the source of dust is conservatively assumed to be the host rock (Assumption 3.6), which is subject to long-term neutron exposure resulting in saturated radioactivity. The scope of this calculation is limited to releases from activated air and silica dust only, excluding natural radioactive releases such as radon or releases from defective waste packages (breached or contaminated). This work supports the repository ventilation system design and Preclosure Safety Analysis. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Waste Package (CRWMS M&O [Civilian Radioactive Waste Management and Operation Contractor] 1999a, page 7). Therefore, this calculation is subject to the requirements of the ''Quality Assurance Requirements and Description'' (DOE [U.S. Department of Energy] 2003). The performance of the calculation and development of this document are carried out in accordance with AP-3.12Q, ''Design Calculation and Analyses'' and LP-3.30Q-BSC, ''Hazards Analysis System''.

  15. Thermal Evaluation for the Naval SNF Waste Package

    SciTech Connect (OSTI)

    T.L. Mitchell

    2000-04-25

    The purpose of this calculation is to evaluate the thermal performance of the naval long spent nuclear fuel (SNF) waste package (WP) under multiple disposal conditions in a monitored geologic repository (MGR). The scope of this calculation is limited to determination of thermal temperature profiles upon the surface of, and within, the naval long SNF WP. The objective is to develop a temperature profile history within the WP, at time increments up to 10,000 years of emplacement. The results of this calculation are intended to support the Naval SNF WP Analysis and Model Report (AMR) for Site Recommendation (SR). This calculation was performed to the specifications within its Technical Development Plan (TDP) (Ref. 8.16). This calculation is developed and documented in accordance with the AP-3.12Q/REV. 0IICN. 0 procedure, Calculations.

  16. Identification of Aircraft Hazards

    SciTech Connect (OSTI)

    K. Ashley

    2006-12-08

    Aircraft hazards were determined to be potentially applicable to a repository at Yucca Mountain in ''Monitored Geological Repository External Events Hazards Screening Analysis'' (BSC 2005 [DIRS 174235], Section 6.4.1). That determination was conservatively based upon limited knowledge of flight data in the area of concern and upon crash data for aircraft of the type flying near Yucca Mountain. The purpose of this report is to identify specific aircraft hazards that may be applicable to a monitored geologic repository (MGR) at Yucca Mountain, using NUREG-0800, ''Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants'' (NRC 1987 [DIRS 103124], Section 3.5.1.6), as guidance for the inclusion or exclusion of identified aircraft hazards. The intended use of this report is to provide inputs for further screening and analysis of identified aircraft hazards based upon the criteria that apply to Category 1 and Category 2 event sequence analyses as defined in 10 CFR 63.2 [DIRS 176544] (Section 4). The scope of this report includes the evaluation of military, private, and commercial use of airspace in the 100-mile regional setting of the repository at Yucca Mountain with the potential for reducing the regional setting to a more manageable size after consideration of applicable screening criteria (Section 7).

  17. Crystal Structures of the Reduced, Sulfenic Acid, and Mixed Disulfide Forms of SarZ, a Redox Active Global Regulator in Staphylococcus aureus

    SciTech Connect (OSTI)

    Poor, Catherine B.; Chen, Peng R.; Duguid, Erica; Rice, Phoebe A.; He, Chuan

    2010-01-20

    SarZ is a global transcriptional regulator that uses a single cysteine residue, Cys{sup 13}, to sense peroxide stress and control metabolic switching and virulence in Staphylococcus aureus. SarZ belongs to the single-cysteine class of OhrR-MgrA proteins that play key roles in oxidative resistance and virulence regulation in various bacteria. We present the crystal structures of the reduced form, sulfenic acid form, and mixed disulfide form of SarZ. Both the sulfenic acid and mixed disulfide forms are structurally characterized for the first time for this class of proteins. The Cys{sup 13} sulfenic acid modification is stabilized through two hydrogen bonds with surrounding residues, and the overall DNA-binding conformation is retained. A further reaction of the Cys{sup 13} sulfenic acid with an external thiol leads to formation of a mixed disulfide bond, which results in an allosteric change in the DNA-binding domains, disrupting DNA binding. Thus, the crystal structures of SarZ in three different states provide molecular level pictures delineating the mechanism by which this class of redox active regulators undergoes activation. These structures help to understand redox-mediated virulence regulation in S. aureus and activation of the MarR family proteins in general.

  18. Abstraction of Models for Pitting and Crevice Corrosion of Drip Shield and Waste Package Outer Barrier

    SciTech Connect (OSTI)

    K. Mon

    2001-08-29

    This analyses and models report (AMR) was conducted in response to written work direction (CRWMS M and O 1999a). ICN 01 of this AMR was developed following guidelines provided in TWP-MGR-MD-000004 REV 01, ''Technical Work Plan for: Integrated Management of Technical Product Input Department'' (BSC 2001, Addendum B). The purpose and scope of this AMR is to review and analyze upstream process-level models (CRWMS M and O 2000a and CRWMS M and O 2000b) and information relevant to pitting and crevice corrosion degradation of waste package outer barrier (Alloy 22) and drip shield (Titanium Grade 7) materials, and to develop abstractions of the important processes in a form that is suitable for input to the WAPDEG analysis for long-term degradation of waste package outer barrier and drip shield in the repository. The abstraction is developed in a manner that ensures consistency with the process-level models and information and captures the essential behavior of the processes represented. Also considered in the model abstraction are the probably range of exposure conditions in emplacement drifts and local exposure conditions on drip shield and waste package surfaces. The approach, method, and assumptions that are employed in the model abstraction are documented and justified.

  19. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    SciTech Connect (OSTI)

    N. E. Pettit

    2001-07-13

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident.

  20. EXTERNAL CRITICALITY CALCULATION FOR DOE SNF CODISPOSAL WASTE PACKAGES

    SciTech Connect (OSTI)

    H. Radulescu

    2002-10-18

    The purpose of this document is to evaluate the potential for criticality for the fissile material that could accumulate in the near-field (invert) and in the far-field (host rock) beneath the U.S. Department of Energy (DOE) spent nuclear fuel (SNF) codisposal waste packages (WPs) as they degrade in the proposed monitored geologic repository at Yucca Mountain. The scope of this calculation is limited to the following DOE SNF types: Shippingport Pressurized Water Reactor (PWR), Enrico Fermi, Fast Flux Test Facility (FFTF), Fort St. Vrain, Melt and Dilute, Shippingport Light Water Breeder Reactor (LWBR), N-Reactor, and Training, Research, Isotope, General Atomics reactor (TRIGA). The results of this calculation are intended to be used for estimating the probability of criticality in the near-field and in the far-field. There are no limitations on use of the results of this calculation. The calculation is associated with the waste package design and was developed in accordance with the technical work plan, ''Technical Work Plan for: Department of Energy Spent Nuclear Fuel and Plutonium Disposition Work Packages'' (Bechtel SAIC Company, LLC [BSC], 2002a). This calculation is subject to the Quality Assurance Requirements and Description (QARD) per the activity evaluation under work package number P6212310Ml in the technical work plan TWP-MGR-MD-0000 10 REV 01 (BSC 2002a).

  1. Evaluation of codisposal viability of MOX (FFTF) DOE-owned fuel: Phase 2 -- Degraded mode calculations

    SciTech Connect (OSTI)

    Goluoglu, S.; Angers, L.; Davis, J.W.; Stockman, H.; Gottlieb, P.; Montierth, L.M.

    1999-07-01

    The authors provide the degraded criticality information that supports the disposal of spent nuclear fuel (SNF) from the US Department of Energy's (DOE's) Fast Flux Test facility (FFTF) in the potential Monitored Geologic Repository (MGR) at Yucca Mountain. FFTF is one of more than 250 forms of DOE-owned SNF. Because of the variety of the SNF, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The FFTF fuel is a mixture of uranium and plutonium oxides and is representative of the mixed-oxide fuel (MOX) group. The analyses were performed according to the disposal criticality analysis methodology that was documented in the topical report submitted to the US nuclear Regulatory Commission (YMP/TR-004Q). The methodology includes analyzing the geochemical and physical processes that can breach the waste package and degrade the waste forms. This paper summarizes the results of geochemistry degradation analysis and the criticality calculations using the degradation products.

  2. CRITICALITY CALCULATION FOR THE MOST REACTIVE DEGRADED CONFIGURATIONS OF THE FFTF SNF CODISPOSAL WP CONTAINING AN INTACT IDENT-69 CONTAINER

    SciTech Connect (OSTI)

    D.R. Moscalu

    2002-08-28

    The objective of this calculation is to perform additional degraded mode criticality evaluations of the Department of Energy's (DOE) Fast Flux Test Facility (FFTF) Spent Nuclear Fuel (SNF) codisposed in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP). The scope of this calculation is limited to the most reactive degraded configurations of the codisposal WP with an almost intact Ident-69 container (breached and flooded but otherwise non-degraded) containing intact FFTF SNF pins. The configurations have been identified in a previous analysis (CRWMS M&O 1999a) and the present evaluations include additional relevant information that was left out of the original calculations. The additional information describes the exact distribution of fissile material in each container (DOE 2002a). The effects of the changes that have been included in the baseline design of the codisposal WP (CRWMS M&O 2000) are also investigated. The calculation determines the effective neutron multiplication factor (k{sub eff}) for selected degraded mode internal configurations of the codisposal waste package. These calculations will support the demonstration of the technical viability of the design solution adopted for disposing of MOX (FFTF) spent nuclear fuel in the potential repository. This calculation is subject to the Quality Assurance Requirements and Description (QARD) (DOE 2002b) per the activity evaluation under work package number P6212310M2 in the technical work plan TWP-MGR-MD-000010 REV 01 (BSC 2002).

  3. Lower-Temperature Subsurface Layout and Ventilation Concepts

    SciTech Connect (OSTI)

    Christine L. Linden; Edward G. Thomas

    2001-06-20

    This analysis combines work scope identified as subsurface facility (SSF) low temperature (LT) Facilities System and SSF LT Ventilation System in the Technical Work Plan for Subsurface Design Section FY 01 Work Activities (CRWMS M&O 2001b, pp. 6 and 7, and pp. 13 and 14). In accordance with this technical work plan (TWP), this analysis is performed using AP-3.10Q, Analyses and Models. It also incorporates the procedure AP-SI.1Q, Software Management. The purpose of this analysis is to develop an overall subsurface layout system and the overall ventilation system concepts that address a lower-temperature operating mode for the Monitored Geologic Repository (MGR). The objective of this analysis is to provide a technical design product that supports the lower-temperature operating mode concept for the revision of the system description documents and to provide a basis for the system description document design descriptions. The overall subsurface layout analysis develops and describes the overall subsurface layout, including performance confirmation facilities (also referred to as Test and Evaluation Facilities) for the Site Recommendation design. This analysis also incorporates current program directives for thermal management.

  4. Secondary Low-Level Waste Treatment Strategy Analysis

    SciTech Connect (OSTI)

    D.M. LaRue

    1999-05-25

    The objective of this analysis is to identify and review potential options for processing and disposing of the secondary low-level waste (LLW) that will be generated through operation of the Monitored Geologic Repository (MGR). An estimate of annual secondary LLW is generated utilizing the mechanism established in ''Secondary Waste Treatment Analysis'' (Reference 8.1) and ''Secondary Low-Level Waste Generation Rate Analysis'' (Reference 8.5). The secondary LLW quantities are based on the spent fuel and high-level waste (HLW) arrival schedule as defined in the ''Controlled Design Assumptions Document'' (CDA) (Reference 8.6). This analysis presents estimates of the quantities of LLW in its various forms. A review of applicable laws, codes, and standards is discussed, and a synopsis of those applicable laws, codes, and standards and their impacts on potential processing and disposal options is presented. The analysis identifies viable processing/disposal options in light of the existing laws, codes, and standards, and then evaluates these options in regard to: (1) Process and equipment requirements; (2) LLW disposal volumes; and (3) Facility requirements.

  5. REVIEW OF NRC APPROVED DIGITAL CONTROL SYSTEMS ANALYSIS

    SciTech Connect (OSTI)

    D.W. Markman

    1999-09-17

    Preliminary design concepts for the proposed Subsurface Repository at Yucca Mountain indicate extensive reliance on modern, computer-based, digital control technologies. The purpose of this analysis is to investigate the degree to which the U. S. Nuclear Regulatory Commission (NRC) has accepted and approved the use of digital control technology for safety-related applications within the nuclear power industry. This analysis reviews cases of existing digitally-based control systems that have been approved by the NRC. These cases can serve as precedence for using similar types of digitally-based control technologies within the Subsurface Repository. While it is anticipated that the Yucca Mountain Project (YMP) will not contain control systems as complex as those required for a nuclear power plant, the review of these existing NRC approved applications will provide the YMP with valuable insight into the NRCs review process and design expectations for safety-related digital control systems. According to the YMP Compliance Program Guidance, portions of various NUREGS, Regulatory Guidelines, and nuclear IEEE standards the nuclear power plant safety related concept would be applied to some of the designs on a case-by-case basis. This analysis will consider key design methods, capabilities, successes, and important limitations or problems of selected control systems that have been approved for use in the Nuclear Power industry. An additional purpose of this analysis is to provide background information in support of further development of design criteria for the YMP. The scope and primary objectives of this analysis are to: (1) Identify and research the extent and precedence of digital control and remotely operated systems approved by the NRC for the nuclear power industry. Help provide a basis for using and relying on digital technologies for nuclear related safety critical applications. (2) Identify the basic control architecture and methods of key digital control systems approved for use in the nuclear power industry by the NRC. (3) Identify and discuss key design issues, features, benefits, and limitations of these NRC approved digital control systems that can be applied as design guidance and correlated to the Monitored Geologic Repository (MGR) design requirements. (4) Identify codes and standards used in the design of these NRC approved digital control systems and discuss their possible applicability to the design of a subsurface nuclear waste repository. (5) Evaluate the NRC approved digital control system's safety, reliability and maintainability features and issues. Apply these to MGR design methodologies and requirements. (6) Provide recommendations for use in developing design criteria in the System Description Documents for the digital control systems of the subsurface nuclear waste repository at Yucca Mountain. (7) Develop recommendations for applying NRC approval methods for digital control systems for the subsurface nuclear waste repository at Yucca Mountain. This analysis will focus on the development of the issues, criteria and methods used and required for identifying the appropriate requirements for digital based control systems. Attention will be placed on development of recommended design criteria for digital controls including interpretation of codes, standards and regulations. Attention will also focus on the use of digital controls and COTS (Commercial Off-the-shelf) technology and equipment in selected NRC approved digital control systems, and as referenced in applicable codes, standards and regulations. The analysis will address design issues related to COTS technology and how they were dealt with in previous NRC approved digital control systems.

  6. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    SciTech Connect (OSTI)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  7. Naval Spent Nuclear Fuel disposal Container System Description Document

    SciTech Connect (OSTI)

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  8. AGING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2004-09-10

    The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the criticality safety results to support the preliminary design of the Aging Facility. As the ongoing design evolution remains fluid, the results from this design calculation should be evaluated for applicability to any new or modified design. Consequently, the results presented in this document are limited to the current design. The information contained in this document was developed by Environmental and Nuclear Engineering and is intended for the use of Design and Engineering in its work regarding the various criticality related activities performed in the Aging Facility. Yucca Mountain Project personnel from Environmental and Nuclear Engineering should be consulted before the use of the information for purposes other than those stated herein or use by individuals other than authorized personnel in Design and Engineering.

  9. FUEL HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2005-06-30

    The purpose of this design calculation is to perform a criticality evaluation of the Fuel Handling Facility (FHF) and the operations and processes performed therein. The current intent of the FHF is to receive transportation casks whose contents will be unloaded and transferred to waste packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, the WPs will also be prepared in the FHF for transfer to the sub-surface facility (for disposal). The MSCs will be transferred to the Aging Facility for storage. The criticality evaluation of the FHF features the following: (I) Consider the types of waste to be received in the FHF as specified below: (1) Uncanistered commercial spent nuclear fuel (CSNF); (2) Canistered CSNF (with the exception of horizontal dual-purpose canister (DPC) and/or multi-purpose canisters (MPCs)); (3) Navy canistered SNF (long and short); (4) Department of Energy (DOE) canistered high-level waste (HLW); and (5) DOE canistered SNF (with the exception of MCOs). (II) Evaluate the criticality analyses previously performed for the existing Nuclear Regulatory Commission (NRC)-certified transportation casks (under 10 CFR 71) to be received in the FHF to ensure that these analyses address all FHF conditions including normal operations, and Category 1 and 2 event sequences. (III) Evaluate FHF criticality conditions resulting from various Category 1 and 2 event sequences. Note that there are currently no Category 1 and 2 event sequences identified for FHF. Consequently, potential hazards from a criticality point of view will be considered as identified in the ''Internal Hazards Analysis for License Application'' document (BSC 2004c, Section 6.6.4). (IV) Assess effects of potential moderator intrusion into the fuel transfer bay for defense in depth. The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that is being carried out in the FHF has been classified as safety category in the ''Q-list'' (BSC 2003, p. A-6). Therefore, this design calculation is subject to the requirements of the ''Quality Assurance Requirements and Description'' (DOE 2004), even though the FHF itself has not yet been classified in the Q-list. Performance of the work scope as described and development of the associated technical product conform to the procedure AP-3.124, ''Design Calculations and Analyses''.

  10. Subsurface Contamination Control

    SciTech Connect (OSTI)

    Y. Yuan

    2001-12-12

    There are two objectives of this report, ''Subsurface Contamination Control''. The first is to provide a technical basis for recommending limiting radioactive contamination levels (LRCL) on the external surfaces of waste packages (WP) for acceptance into the subsurface repository. The second is to provide an evaluation of the magnitude of potential releases from a defective WP and the detectability of the released contents. The technical basis for deriving LRCL has been established in ''Retrieval Equipment and Strategy for Wp on Pallet'' (CRWMS M and O 2000g, 6.3.1). This report updates the derivation by incorporating the latest design information of the subsurface repository for site recommendation. The derived LRCL on the external surface of WPs, therefore, supercede that described in CRWMS M and O 2000g. The derived LRCL represent the average concentrations of contamination on the external surfaces of each WP that must not be exceeded before the WP is to be transported to the subsurface facility for emplacement. The evaluation of potential releases is necessary to control the potential contamination of the subsurface repository and to detect prematurely failed WPs. The detection of failed WPs is required in order to provide reasonable assurance that the integrity of each WP is intact prior to MGR closure. An emplaced WP may become breached due to manufacturing defects or improper weld combined with failure to detect the defect, by corrosion, or by mechanical penetration due to accidents or rockfall conditions. The breached WP may release its gaseous and volatile radionuclide content to the subsurface environment and result in contaminating the subsurface facility. The scope of this analysis is limited to radioactive contaminants resulting from breached WPs during the preclosure period of the subsurface repository. This report: (1) documents a method for deriving LRCL on the external surfaces of WP for acceptance into the subsurface repository; (2) provides a table of derived LRCL for nuclides of radiological importance; (3) Provides an as low as is reasonably achievable (ALARA) evaluation of the derived LRCL by comparing potential onsite and offsite doses to documented ALARA requirements; (4) Provides a method for estimating potential releases from a defective WP; (5) Provides an evaluation of potential radioactive releases from a defective WP that may become airborne and result in contamination of the subsurface facility; and (6) Provides a preliminary analysis of the detectability of a potential WP leak to support the design of an airborne release monitoring system.

  11. Defense High Level Waste Disposal Container System Description

    SciTech Connect (OSTI)

    2000-10-12

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials will be selected for the disposal container inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lids will be a barrier made of high-nickel alloy. The defense HLW disposal container interfaces with the emplacement drift environment and the internal waste by transferring heat from the canisters to the external environment and by protecting the canisters and their contents from damage/degradation by the external environment. The disposal container also interfaces with the canisters by limiting access of moderator and oxidizing agents to the waste. A loaded and sealed disposal container (waste package) interfaces with the Emplacement Drift System's emplacement drift waste package supports upon which the waste packages are placed. The disposal container interfaces with the Canister Transfer System, Waste Emplacement /Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement, and retrieval for the disposal container/waste package.

  12. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    SciTech Connect (OSTI)

    2000-10-12

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in the emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Multiple boiling water reactor (BWR) and pressurized water reactor (PWR) disposal container designs are needed to accommodate the expected range of spent fuel assemblies and provide long-term confinement of the commercial SNF. The disposal container will include outer and inner cylinder walls, outer cylinder lids (two on the top, one on the bottom), inner cylinder lids (one on the top, one on the bottom), and an internal metallic basket structure. Exterior labels will provide a means by which to identify the disposal container and its contents. The two metal cylinders, in combination with the cladding, Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lid will be made of high-nickel alloy. The basket will assist criticality control, provide structural support, and improve heat transfer. The Uncanistered SNF Disposal Container System interfaces with the emplacement drift environment and internal waste by transferring heat from the SNF to the external environment and by protecting the SFN assemblies and their contents from damage/degradation by the external environment. The system also interfaces with the SFN by limiting access of moderator and oxidizing agents of the SFN. The waste package interfaces with the Emplacement Drift System's emplacement drift pallets upon which the wasted packages are placed. The disposal container interfaces with the Assembly Transfer System, Waste Emplacement/Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and retrieval of the disposal container/waste package.

  13. DOE Hydropower Program Biennial Report for FY 2005-2006

    SciTech Connect (OSTI)

    Sale, Michael J; Cada, Glenn F; Acker, Thomas L.; Carlson, Thomas; Dauble, Dennis D.; Hall, Douglas G.

    2006-07-01

    SUMMARY The U.S. Department of Energy (DOE) Hydropower Program is part of the Office of Wind and Hydropower Technologies, Office of Energy Efficiency and Renewable Energy. The Program's mission is to conduct research and development (R&D) that will increase the technical, societal, and environmental benefits of hydropower. The Department's Hydropower Program activities are conducted by its national laboratories: Idaho National Laboratory (INL) [formerly Idaho National Engineering and Environmental Laboratory], Oak Ridge National Laboratory (ORNL), Pacific Northwest National Laboratory (PNNL), and National Renewable Energy Laboratory (NREL), and by a number of industry, university, and federal research facilities. Programmatically, DOE Hydropower Program R&D activities are conducted in two areas: Technology Viability and Technology Application. The Technology Viability area has two components: (1) Advanced Hydropower Technology (Large Turbine Field Testing, Water Use Optimization, and Improved Mitigation Practices) and (2) Supporting Research and Testing (Environmental Performance Testing Methods, Computational and Physical Modeling, Instrumentation and Controls, and Environmental Analysis). The Technology Application area also has two components: (1) Systems Integration and Technology Acceptance (Hydro/Wind Integration, National Hydropower Collaborative, and Integration and Communications) and (2) Supporting Engineering and Analysis (Valuation Methods and Assessments and Characterization of Innovative Technology). This report describes the progress of the R&D conducted in FY 2005-2006 under all four program areas. Major accomplishments include the following: Conducted field testing of a Retrofit Aeration System to increase the dissolved oxygen content of water discharged from the turbines of the Osage Project in Missouri. Contributed to the installation and field testing of an advanced, minimum gap runner turbine at the Wanapum Dam project in Washington. Completed a state-of-the-science review of hydropower optimization methods and published reports on alternative operating strategies and opportunities for spill reduction. Carried out feasibility studies of new environmental performance measurements of the new MGR turbine at Wanapum Dam, including measurement of behavioral responses, biomarkers, bioindex testing, and the use of dyes to assess external injuries. Evaluated the benefits of mitigation measures for instream flow releases and the value of surface flow outlets for downstream fish passage. Refined turbulence flow measurement techniques, the computational modeling of unsteady flows, and models of blade strike of fish. Published numerous technical reports, proceedings papers, and peer-reviewed literature, most of which are available on the DOE Hydropower website. Further developed and tested the sensor fish measuring device at hydropower plants in the Columbia River. Data from the sensor fish are coupled with a computational model to yield a more detailed assessment of hydraulic environments in and around dams. Published reports related to the Virtual Hydropower Prospector and the assessment of water energy resources in the U.S. for low head/low power hydroelectric plants. Convened a workshop to consider the environmental and technical issues associated with new hydrokinetic and wave energy technologies. Laboratory and DOE staff participated in numerous workshops, conferences, coordination meetings, planning meetings, implementation meetings, and reviews to transfer the results of DOE-sponsored research to end-users.

  14. Peak Ground Velocities for Seismic Events at Yucca Mountain, Nevada

    SciTech Connect (OSTI)

    K. Coppersmith; R. Quittmeyer

    2005-02-16

    This report describes a scientific analysis to bound credible horizontal peak ground velocities (PGV) for the repository waste emplacement level at Yucca Mountain. Results are presented as a probability distribution for horizontal PGV to represent uncertainties in the analysis. The analysis also combines the bound to horizontal PGV with results of ground motion site-response modeling (BSC 2004 [DIRS 170027]) to develop a composite hazard curve for horizontal PGV at the waste emplacement level. This result provides input to an abstraction of seismic consequences (BSC 2004 [DIRS 169183]). The seismic consequence abstraction, in turn, defines the input data and computational algorithms for the seismic scenario class of the total system performance assessment (TSPA). Planning for the analysis is documented in Technical Work Plan TWP-MGR-GS-000001 (BSC 2004 [DIRS 171850]). The bound on horizontal PGV at the repository waste emplacement level developed in this analysis complements ground motions developed on the basis of PSHA results. In the PSHA, ground motion experts characterized the epistemic uncertainty and aleatory variability in their ground motion interpretations. To characterize the aleatory variability they used unbounded lognormal distributions. As a consequence of these characterizations, as seismic hazard calculations are extended to lower and lower annual frequencies of being exceeded, the ground motion level increases without bound, eventually reaching levels that are not credible (Corradini 2003 [DIRS 171191]). To provide credible seismic inputs for TSPA, in accordance with 10 Code of Federal Regulations (CFR) 63.102(j) [DIRS 156605], this complementary analysis is carried out to determine reasonable bounding values of horizontal PGV at the waste emplacement level for annual frequencies of exceedance as low as 10{sup -8}. For each realization of the TSPA seismic scenario, the results of this analysis provide a constraint on the values sampled from the horizontal PGV hazard curve for the waste emplacement level. The relation of this analysis to other work feeding the seismic consequence abstraction and the TSPA is shown on Figure 1-1. The ground motion hazard results from the PSHA provide the basis for inputs to a site-response model that determines the effect of site materials on the ground motion at a location of interest (e.g., the waste emplacement level). Peak ground velocity values determined from the site-response model for the waste emplacement level are then used to develop time histories (seismograms) that form input to a model of drift degradation under seismic loads potentially producing rockfall. The time histories are also used to carry out dynamic seismic structural response calculations of the drip shield and waste package system. For the drip shield, damage from seismically induced rockfall also is considered. In the seismic consequence abstraction, residual stress results from the structural response calculations are interpreted in terms of the percentage of the component (drip shield, waste package) damaged as a function of horizontal PGV. The composite hazard curve developed in this analysis, which reflects the results of site-response modeling and the bound to credible horizontal PGV at the waste emplacement level, also feeds the seismic consequence abstraction. The composite hazard curve is incorporated into the TSPA sampling process to bound horizontal PGV and related seismic consequences to values that are credible.

  15. Waste Emplacement/Retrieval System Description Document

    SciTech Connect (OSTI)

    Eric Loros

    2001-07-25

    The Waste Emplacement/Retrieval System transports Waste Packages (WPs) from the Waste Handling Building (WHB) to the subsurface area of emplacement, and emplaces the WPs once there. The Waste Emplacement/Retrieval System also, if necessary, removes some or all of the WPs from the underground and transports them to the surface. Lastly, the system is designed to remediate abnormal events involving the portions of the system supporting emplacement or retrieval. During emplacement operations, the system operates on the surface between the WHB and North Portal, and in the subsurface in the North Ramp, access mains, and emplacement drifts. During retrieval or abnormal conditions, the operations areas may also extend to a surface retrieval storage site and South Portal on the surface, and the South Ramp in the subsurface. A typical transport and emplacement operation involves the following sequence of events. A WP is loaded into a WP transporter at the WHB, and coupled to a pair of transport locomotives. The locomotives transport the WP from the WHB, down the North Ramp, and to the entrance of an emplacement drift. Once docked at the entrance of the emplacement drift, the WP is moved outside of the WP transporter, and engaged by a WP emplacement gantry. The WP emplacement gantry lifts the WP, and transports it to its emplacement location, where the WP is then lowered to its final resting position. The WP emplacement gantry remains in the drift while the WP transporter is returned to the WHB by the locomotives. When the transporter reaches the WHB, the sequence of operations is repeated. Retrieval of all the WPs, or a large group of WPs, under normal conditions is achieved by reversing the emplacement operations. Retrieval of a small set of WPs, under normal or abnormal conditions, is known as recovery. Recovery performed under abnormal conditions will involve a suite of specialized equipment designed to perform a variety of tasks to enable the recovery process. Recovery after abnormal events may require clearing of equipment, rock, and ground support to facilitate recovery operations. Stabilization of existing ground support and installation of new ground support may also be needed. Recovery of WP(s) after an event that has contaminated drifts and/or WPs will require limiting the spread of contamination. Specialized equipment will also be necessary for system restoration (e.g., after a derailment, component failure). The Waste Emplacement/Retrieval System interfaces with the Subsurface Facility System and Ground Control System for the size and layout of the underground openings. The system interfaces with the Subsurface Ventilation System for the emplacement drift operating environment and the size of the drift isolation doors. The system interfaces with all WP types for the size, weight, and other important parameters affecting emplacement, recovery, and retrieval. The system interfaces with the Subsurface Emplacement Transportation System for the rail system upon which it operates and the distribution of power through the rail system. The system interfaces with the Monitored Geologic Repository (MGR) Operations Monitoring and Control System for the transmission of data to and from the system equipment, and for remote control of system equipment. The system interfaces with the Ground Control System for any repairs that are made. The system interfaces with the Emplacement Drift System for the WP emplacement mode and hardware. The system interfaces with the Disposal Container Handling System and the Waste Handling Building System for the receipt (during emplacement) and delivery (during retrieval/recovery) of WPs.

  16. Waste Emplacement/Retrieval System Description Document

    SciTech Connect (OSTI)

    2000-10-12

    The Waste Emplacement/Retrieval System transports Waste Packages (WPs) from the Waste Handling Building (WHB) to the subsurface area of emplacement, and emplaces the WPs once there. The system also, if necessary, removes some or all of the WPs from the underground and transports them to the surface. Lastly, the system is designed to remediate abnormal events involving the portions of the system supporting emplacement or retrieval. During emplacement operations, the system operates on the surface between the WHB and North Portal, and in the subsurface in the North Ramp, access mains, and emplacement drifts. During retrieval or abnormal conditions, the operations areas may also extend to a surface retrieval storage site and South Portal on the surface, and the South Ramp in the subsurface. A typical transport and emplacement operation involves the following sequence of events. A WP is loaded into a WP transporter at the WHB, and coupled to a pair of transport locomotives. The locomotives transport the WP from the WHB, down the North Ramp, and to the entrance of an emplacement drift. Once docked at the entrance of the emplacment drift, the WP is moved outside of the WP transporter, and engaged by a WP emplacement gantry. The gantry lifts the WP, and transports it to its emplacement location, where the WP is then lowered to its final resting position. The gantry remains in the drift while the WP transporter is returned to the WHB by the locomotives. When the transporter reaches the WHB, the sequence of operations is repeated. Retrieval of all the WPs, or a large group of WPs, under normal conditions is achieved by reversing the emplacement operations. Retrieval of a small set of WPs, under normal or abnormal conditions, is known as recovery. Recovery performed under abnormal conditions will involve a suite of specialized equipment designed to perform a variety of tasks to enable the recovery process. Recovery after abnormal events may require clearing of equipment, rock, and ground support to facilitate recovery operations. Stabilization of existing ground support and installation of new ground support may also be needed. Recovery of WPs after an event that has contaminated drifts and/or WPs will require limiting the spread of contamination. Specialized equipment will also be necessary for system restoration. The system interfaces with the Subsurface Facility System and Ground Control System for the size and layout of the underground openings. The system interfaces with the Subsurface Ventilation System for the emplacement drift operating environment and the size of the drift isolation doors. The system interfaces with all WP types for the size, weight, and other important parameters affecting emplacement, recovery, and retrieval. The system interfaces with the Subsurface Emplacement Transportation System for the rail system upon which it operates and the distribution of power throuch the rail system. The system interfaces with the Monitored Geologic Repository (MGR) Operations Monitoring and Control System for the transmission of data to and from the system equipment, and for remote control of system equipment. The system interfaces with the Ground Control System for any repairs that are made. The system interfaces with the Emplacement Drift System for the WP emplacement mode and hardware. The system interfaces with the Disposal Container Handling System and the Waste Handling Building System for the receipt (during emplacement) and delivery (during retrieval/recovery) of WPs.

  17. Monitored Geologic Repository Operations Monitoring and Control System Description Document

    SciTech Connect (OSTI)

    E.F. Loros

    2000-06-29

    The Monitored Geologic Repository Operations Monitoring and Control System provides supervisory control, monitoring, and selected remote control of primary and secondary repository operations. Primary repository operations consist of both surface and subsurface activities relating to high-level waste receipt, preparation, and emplacement. Secondary repository operations consist of support operations for waste handling and treatment, utilities, subsurface construction, and other selected ancillary activities. Remote control of the subsurface emplacement operations, as well as, repository performance confirmation operations are the direct responsibility of the system. In addition, the system monitors parameters such as radiological data, air quality data, fire detection status, meteorological conditions, unauthorized access, and abnormal operating conditions, to ensure a safe workplace for personnel. Parameters are displayed in a real-time manner to human operators regarding surface and subsurface conditions. The system performs supervisory monitoring and control for both important to safety and non-safety systems. The system provides repository operational information, alarm capability, and human operator response messages during emergency response situations. The system also includes logic control to place equipment, systems, and utilities in a safe operational mode or complete shutdown during emergency response situations. The system initiates alarms and provides operational data to enable appropriate actions at the local level in support of emergency response, radiological protection response, evacuation, and underground rescue. The system provides data communications, data processing, managerial reports, data storage, and data analysis. This system's primary surface and subsurface operator consoles, for both supervisory and remote control activities, will be located in a Central Control Center (CCC) inside one of the surface facility buildings. The system consists of instrument and control equipment and components necessary to provide human operators with sufficient information to monitor and control the operation of the repository in an efficient and safe manner. The system consists of operator consoles and workstations, multiple video display terminals, communications and interfacing equipment, and instrument and control software with customized configuration to meet the needs of the Monitored Geologic Repository (MGR). Process and logic controllers and the associated input/output units of each system interfaced with this system will be configured into Remote Terminal Units (RTU) and located close to the systems to be monitored and controlled. The RTUs are configured to remain operational should communication with CCC operations be lost. The system provides closed circuit television to selectively view systems, operations, and equipment areas and to aid in the operation of mechanical systems. Control and monitoring of site utility systems will be located in the CCC. Site utilities include heating, ventilation, and air conditioning equipment; plant compressed air; plant water; firewater; electrical systems; and inert gases, such as nitrogen, if required. This system interfaces with surface and subsurface systems that either generate output data or require remote control input. The system interfaces with the Site Communications System for bulk storage of operational data, on-site and off-site communication, and a plant-wide public announcement system. The system interfaces with the Safeguards and Security System to provide operational status and emergency alarm indications. The system interfaces with the Site Operation System to provide site wide acquisition of data for analysis and reports, historical information for trends, utility information for plant operation, and to receive operating plans and procedures.