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1

CLASSIFICATION OF THE MGR SURFACE ENVIRONMENTAL MONITORING SYSTEM  

SciTech Connect (OSTI)

The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) surface environmental monitoring 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-0333PY ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998).

J.A. Ziegler

1999-08-31T23:59:59.000Z

2

CLASSIFICATION OF THE MGR SUBSURFACE DEVELOPMENT TRANSPORTATION SYSTEM  

SciTech Connect (OSTI)

The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) subsurface development transportation 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-0333P7 ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998).

R. Garrett

1999-08-31T23:59:59.000Z

3

PRELIMINARY SELECTION OF MGR DESIGN BASIS EVENTS  

SciTech Connect (OSTI)

The purpose of this analysis is to identify the preliminary design basis events (DBEs) for consideration in the design of the Monitored Geologic Repository (MGR). For external events and natural phenomena (e.g., earthquake), the objective is to identify those initiating events that the MGR will be designed to withstand. Design criteria will ensure that radiological release scenarios resulting from these initiating events are beyond design basis (i.e., have a scenario frequency less than once per million years). For internal (i.e., human-induced and random equipment failures) events, the objective is to identify credible event sequences that result in bounding radiological releases. These sequences will be used to establish the design basis criteria for MGR structures, systems, and components (SSCs) design basis criteria in order to prevent or mitigate radiological releases. The safety strategy presented in this analysis for preventing or mitigating DBEs is based on the preclosure safety strategy outlined in ''Strategy to Mitigate Preclosure Offsite Exposure'' (CRWMS M&O 1998f). DBE analysis is necessary to provide feedback and requirements to the design process, and also to demonstrate compliance with proposed 10 CFR 63 (Dyer 1999b) requirements. DBE analysis is also required to identify and classify the SSCs that are important to safety (ITS).

J.A. Kappes

1999-09-16T23:59:59.000Z

4

Comp. Sys. Mgr. 4 Curtis Rias  

E-Print Network [OSTI]

IT Procurement Management Student Technology Fee Coordination Mobile Devices Coordination Site Licensing Coordination Computer Labs Management CUNYfirst Readiness & CUNY Alert Liaison TECHNOLOGY ADVISORY COMMITTEE Project Management Office (PMO) ISS IT Internal Operations Management Office of Information Technology

Brinkmann, Peter

5

Code Coverage-Based Regression Test Selection and Prioritization in WebKit rpd Beszdes, Tams Gergely, Lajos Schrettner, Judit Jsz, Lszl Lang, Tibor Gyimthy  

E-Print Network [OSTI]

. Although the possible benefits are clear, a practical implementation and sustained reliability evolving software system. However, in many cases regression test suites tend to grow too large to be suitable for full re-execution at each change of the software. In this case selective retesting can

Beszedes, Árpád

6

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched Ferromagnetism in Layered NbS2 andThe1A:decisional. 1 B O NAmes Blue Alert-PMTOTitle:

7

Biomechanics of common carotid arteries from mice heterozygous for mgR, the most common mouse model of Marfan syndrome  

E-Print Network [OSTI]

Marfan syndrome, affecting approximately one out of every 5,000 people, is characterized by abnormal bone growth, ectopia lentis, and often-fatal aortic dilation and dissection. The root cause is a faulty extracellular matrix protein, fibrillin-1...

Taucer, Anne Irene

2009-05-15T23:59:59.000Z

8

2212 IEEE TRANSACTIONS ON INFORMATION THEORY, VOL. 46, NO. 6, SEPTEMBER 2000 The Super-Trellis Structure of Turbo Codes  

E-Print Network [OSTI]

-Trellis Structure of Turbo Codes Marco Breiling, Student Member, IEEE, and Lajos Hanzo, Senior Member, IEEE Abstract--In this contribution we derive the super-trellis structure of turbo codes. We show that this structure and its associated decoding com- plexity depend strongly on the interleaver applied in the turbo encoder. We provide

Verdú, Sergio

9

International Workshop on Recent Advances in Mathematical Statistics  

E-Print Network [OSTI]

) · Siegfried H¨ormann (Universit´e libre de Bruxelles) · Lajos Horv´ath (University of Utah, Salt Lake City University in Prague) · Gregory Rice (University of Utah, Salt Lake City) · Pranab K. Sen (University of the Charles University in Prague has the pleasure to invite you to participate in the Workshop organized

Cerveny, Vlastislav

10

OORGANIZATIONRGANIZATION CCHARTHART Human Resources  

E-Print Network [OSTI]

Barb Stephens Jeff Butler Recruiting Engineering & Utilities A t Di /Mg Campus Maintenance Supervisor Accounting Associates --Asst. Dir/Mgr--- Dan Stevenson Engineering University Engineer --Supervisor-- Tom

Maxwell, Bruce D.

11

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...

12

Office: ITO PE/Project  

E-Print Network [OSTI]

Mgr.: Mills/Swinson PAD No.: Smart Spaces Moving Through Smart Spaces "city-wide appliances" "in1 DARPA Office: ITO PE/Project: Pgm No.: Pgm Mgr.: Mills/Swinson PAD No.: Smart Spaces Personal Information Projection Develop techniques for projecting personal information from cyberspace into smart

Mills, Kevin

13

Sample and Implied Volatility in GARCH Models  

E-Print Network [OSTI]

Sample and Implied Volatility in GARCH Models Lajos Horva´th University of Utah Piotr Kokoszka Utah of various GARCH-type models is a function hðq? of the parameter vector q which is estimated by bq. For most distributions of the differences ^2 ? hðq? and ^2 ? hðbq? for broad classes of GARCH-type models. Even though

Kokoszka, Piotr

14

Praveen Panchal VP of IT and CIO  

E-Print Network [OSTI]

Technology (OIT) Organization Chart June 13 Mark Kam Deputy CIO Comp. Sys. Mgr. 4 Project Management Office Distribution & Management BARFIT* Coordination Time Keeping IT Procurement Management Student Technology Fee Coordination Mobile Devices Coordination Site Licensing Coordination Computer Labs Management CUNYfirst

Brinkmann, Peter

15

Scott Koenig Development Officer  

E-Print Network [OSTI]

Scott Koenig Development Officer Teri Lucie Thompson Senior Vice President & CMO UA FOUNDAT I/TV Station Mgr Frank Fregoso Chief Engineer Cheech Calenti IT Manager Ed Kesterson Radio Program Dir. AHSC

Utzinger, Urs

16

Chemistry Safety Notes Volume 1, Issue 2 December 2013  

E-Print Network [OSTI]

by the Chemistry Dept. Safety Committee, written & edited by Debbie Decker, Safety Mgr. EH&S Inspections EH extinguisher, there's a green tag with the fire marshal stamp on one side. On the reverse, there's a list

Guo, Ting

17

Electric Utility Measurement & Verification Program  

E-Print Network [OSTI]

Electric Utility Measurement & Verification Program Ken Lau, P.Eng., CMVP Graham Henderson, P.Eng., CMVP Dan Hebert, P.Eng.,CMVP Mgr, Measurement & Verification Engineering Team Leader Senior Engineer BC Hydro Burnaby, BC Canada...

Lau, K.; Henderson, G.; Hebert, D.

18

E032REGISTRATION UTILIDOR BELOW  

E-Print Network [OSTI]

WATER E032REGISTRATION SUPPORT SERV. CONTR. E071 TRASH ROOM VEST E069.2 EXHIBIT HALL 03 E003 CONCESS SUPP. E073 SHOW MGR E074 E075E082 E081 E079 VEST E077 VEST E076 VEST E083 E085 CONCESS. SUPP. E087 SHOW MGR E086 VEST E084 E078 E080 ELEC E070.1 COMM. E070 E026 PE02.1 PE02.B E002.1 ELEC E028.1 FAMILY E

19

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.

20

Praveen Panchal VP of IT and CIO  

E-Print Network [OSTI]

Technology Fee Coordination Mobile Devices Coordination Site Licensing Coordination Computer Labs Management Technology (OIT) Organization Chart April 14 Mark Kam Deputy CIO Comp. Sys. Mgr. 4 Project Management IT Internal Operations Management CETL Office of Information Technology (OIT) Organization Chart April 14

Brinkmann, Peter

Note: This page contains sample records for the topic "mgr lajos grof-tisza" 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

Workshop National Renewable Energy Laboratory DOE Hydrogen ProgramDOE Hydrogen Program  

E-Print Network [OSTI]

of assembling the constituent parts (e.g., subsystems, components) of a system in a logical, cost-effective way Storage Fuel Cells Concept formulation and initial implementation SystemsSystems IntegrationIntegration FY Platform Thermochemical Platform Program Mgr DOE Hydrogen Program Production & Delivery Storage Fuel Cells

22

Energy-Efficient Platform Designs for Real-World Wireless Sensing Applications  

E-Print Network [OSTI]

Unit Scheduler Power Mgr Driver Calib. PowerReg.&Distr.Switch windmill solar panel battery other the trade-offs of sensing devices, wireless interfaces, and computation and control units. We also cover Controller mod Sensor + Detector ADC /det Reference Loc. Sens. Sensing Unit Processing Unit Communication

Shinozuka, Masanobu

23

Particle Physics & Astrophysics David MacFarlane, Director  

E-Print Network [OSTI]

Division Head CDMS Richard Partridge Dept Head KIPAC Computing Stuart Marshall Dept Head DES David Burke Services Mgr. Debbie Tryforos Assistant to the Director Sensors & Detectors Chris Kenney Dept Head Redwood Conference Rooms SLUO Debbie Tryforos Assistant to the Director HEP Faculty Coordiantor Assistant

Wechsler, Risa H.

24

INDUSTRIAL/MILITARY ACTIVITY-INITIATED ACCIDENT SCREENING ANALYSIS  

SciTech Connect (OSTI)

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.

D.A. Kalinich

1999-09-27T23:59:59.000Z

25

Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel  

SciTech Connect (OSTI)

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

L. Angers

2001-01-31T23:59:59.000Z

26

Monitored Geologic Repository Project Description Document  

SciTech Connect (OSTI)

The primary objective of the Monitored Geologic Repository Project Description Document (PDD) is to allocate the functions, requirements, and assumptions to the systems at Level 5 of the Civilian Radioactive Waste Management System (CRWMS) architecture identified in Section 4. It provides traceability of the requirements to those contained in Section 3 of the ''Monitored Geologic Repository Requirements Document'' (MGR RD) (CRWMS M&O 2000b) and other higher-level requirements documents. In addition, the PDD allocates design related assumptions to work products of non-design organizations. The document provides Monitored Geologic Repository (MGR) engineering design basis in support of design and performance assessment in preparing for the Site Recommendation (SR) and License Application (LA) milestones. The engineering design basis documented in the PDD is to be captured in the System Description Documents (SDDs) which address each of the systems at Level 5 of the CRWMS architecture. The design engineers obtain the engineering design basis from the SDDs and by reference from the SDDs to the PDD. The design organizations and other organizations will obtain design related assumptions directly from the PDD. These organizations may establish additional assumptions for their individual activities, but such assumptions are not to conflict with the assumptions in the PDD. The PDD will serve as the primary link between the engineering design basis captured in the SDDs and the design requirements captured in U.S. Department of Energy (DOE) documents. The approved PDD is placed under Level 3 baseline control by the CRWMS Management and Operating Contractor (M&O) and the following portions of the PDD constitute the Technical Design Baseline for the MGR: the design characteristics listed in Table 2-1, the MGR Architecture (Section 4.1),the Engineering Design Bases (Section 5), and the Controlled Project Assumptions (Section 6).

P. Curry

2000-06-01T23:59:59.000Z

27

Monitored Geologic Repository Project Description Document  

SciTech Connect (OSTI)

The primary objective of the Monitored Geologic Repository Project Description Document (PDD) is to allocate the functions, requirements, and assumptions to the systems at Level 5 of the Civilian Radioactive Waste Management System (CRWMS) architecture identified in Section 4. It provides traceability of the requirements to those contained in Section 3 of the ''Monitored Geologic Repository Requirements Document'' (MGR RD) (YMP 2000a) and other higher-level requirements documents. In addition, the PDD allocates design related assumptions to work products of non-design organizations. The document provides Monitored Geologic Repository (MGR) technical requirements in support of design and performance assessment in preparing for the Site Recommendation (SR) and License Application (LA) milestones. The technical requirements documented in the PDD are to be captured in the System Description Documents (SDDs) which address each of the systems at Level 5 of the CRWMS architecture. The design engineers obtain the technical requirements from the SDDs and by reference from the SDDs to the PDD. The design organizations and other organizations will obtain design related assumptions directly from the PDD. These organizations may establish additional assumptions for their individual activities, but such assumptions are not to conflict with the assumptions in the PDD. The PDD will serve as the primary link between the technical requirements captured in the SDDs and the design requirements captured in US Department of Energy (DOE) documents. The approved PDD is placed under Level 3 baseline control by the CRWMS Management and Operating Contractor (M and O) and the following portions of the PDD constitute the Technical Design Baseline for the MGR: the design characteristics listed in Table 1-1, the MGR Architecture (Section 4.1), the Technical Requirements (Section 5), and the Controlled Project Assumptions (Section 6).

P. M. Curry

2001-01-30T23:59:59.000Z

28

Driving Energy Performance with Energy Management Teams  

E-Print Network [OSTI]

Driving Energy Performance with Energy Management Teams Meredith Younghein ENERGY STAR Industrial Communications Mgr. U.S. Environmental Protection Agency Washington, DC ABSTRACT Companies today face an uncertain energy future. Businesses... face escalating energy prices which can erode profits. Concerns over supply reliability, and possible regulation of carbon emissions create risk. For many industries in the U.S., energy costs are equal to the cost of raw materials or even employee...

Younghein, M.; Tunnessen, W.

2006-01-01T23:59:59.000Z

29

Light-Emitting Tag Testing in Conjunction with Testing of the Minimum Gap Runner Turbine Design at Bonneville Dam Powerhouse 1  

SciTech Connect (OSTI)

This report describes a pilot study conducted by Tom Carlson of PNNL and Mark Weiland of MEVATEC Corp to test the feasibility of using light-emitting tags to visually track objects passing through the turbine environment of a hydroelectric dam. Light sticks were released at the blade tip, mid-blade, and hub in the MGR turbine and a Kaplan turbine at Bonneville Dam and videotaped passing thru the dam to determine visibility and object trajectories.

Carlson, Thomas J.; Weiland, Mark A.

2001-01-30T23:59:59.000Z

30

Monitored Geologic Repository Project Description Document  

SciTech Connect (OSTI)

The primary objective of the Monitored Geologic Repository Project Description Document (PDD) is to allocate the functions, requirements, and assumptions to the systems at Level 5 of the Civilian Radioactive Waste Management System (CRWMS) architecture identified in Section 4. It provides traceability of the requirements to those contained in Section 3 of the Yucca Mountain Site Characterization Project Requirements Document (YMP RD) (YMP 2001a) and other higher-level requirements documents. In addition, the PDD allocates design related assumptions to work products of non-design organizations. The document provides Monitored Geologic Repository (MGR) technical requirements in support of design and performance assessment in preparing for the Site Recommendation (SR) and License Application (LA) milestones. The technical requirements documented in the PDD are to be captured in the System Description Documents (SDDs) which address each of the systems at Level 5 of the CRWMS architecture. The design engineers obtain the technical requirements from the SDDs and by reference from the SDDs to the PDD. The design organizations and other organizations will obtain design related assumptions directly from the PDD. These organizations may establish additional assumptions for their individual activities, but such assumptions are not to conflict with the assumptions in the PDD. The PDD will serve as the primary link between the technical requirements captured in the SDDs and the design requirements captured in US Department of Energy (DOE) documents. The approved PDD is placed under Level 3 baseline control by the CRWMS Management and Operating Contractor (M&O) and the following portions of the PDD constitute the Technical Design Baseline for the MGR: the design characteristics listed in Table 1-1, the MGR Architecture (Section 4.1), the Technical Requirements (Section 5), and the Controlled Project Assumptions (Section 6).

P. Curry

2001-06-26T23:59:59.000Z

31

COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS  

SciTech Connect (OSTI)

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.

S.O. Bader

1999-10-18T23:59:59.000Z

32

A study of sterility in certain Gossypium hybrids  

E-Print Network [OSTI]

a i i s f or smxxjTT n cnun ckmsifxum umim A MMMrtafclMi iik irt Jmkia i OahM. jy. )fcj?r MJivfei a? itlaa Aipfwil M to ity&? ~ Mgr, 1952 a a??? or s t to u ty a d r a m oattxvxat mcnxM ?sr... of ow eod ig s te r ility b a r r ie r s .......................... . . . . kk ??????? AMD MBTH0D8................................... 1(6 Cooqpariaan of jollm behavior......... ................ ?? Comparison of fertilisin g ab ilities...

Oakes, Albert Jackson

1952-01-01T23:59:59.000Z

33

Measurement of the work function of filament-evaporated M?g???[?subscript -x]A?u?[?subscript x] films  

E-Print Network [OSTI]

) Roland E. Allen (Member) Mi Lu ( Member) J. W. Howze (Head of Department) December Igg0 ABSTRACT Measurement of the Work Function of Filament-Evaporated Mgr, Au Films. (December 1990) Dong Li, M. S. , Institute of Physics, Chinese Academy... energy hv I = AT'(W, ? hv) '~'f( ) hT (1. 2) where A is a constant independent of v, T and P. W, is the surface potential and and f(x) 1 )u+1 n2 u=r 2 2 cd ? uz ? + ? ? 2 (-I)"" 6 2 ?, nr (z & 0) (& -0) (1. 3) Journal model is IEEE Transactions...

Li, Dong

1990-01-01T23:59:59.000Z

34

General and Localized Corrosion of Borated Stainless Steels  

SciTech Connect (OSTI)

The Transportation, Aging and Disposal (TAD) canister-based system is being proposed to transport and store spent nuclear fuel at the Monitored Geologic Repository (MGR) located at Yucca Mountain, Nevada. The preliminary design of this system identifies borated stainless steel as the neutron absorber material that will be used to fabricate fuel basket inserts for nuclear criticality control. This paper discusses corrosion test results for verifying the performance of this material manufactured to the requirements of ASTM A887, Grade A, under the expected repository conditions.

T.E. Lister; Ronald E. Mizia; A.W. Erickson; T.L. Trowbridge; B. S. Matteson

2008-03-01T23:59:59.000Z

35

Monday Musings, October 22, 2012  

E-Print Network [OSTI]

and organizational review teams. As always, I welcome your feedbacksimply reply to this message with questions or comments. -- STRATEGIC PLANNING Oversight Team Update | Mary Roach (chair) and Aileen Ball (co-chair): On Friday, October 12th, members... Riley ? Brian Rosenblum ? Kathleen Ames-Oliver, HR Mgr Learning & Development (and frequent special guest) -- OA Fund Today, we announced the creation of a $50,000 authors fund to support open access scholarship at KUs Lawrence campus and the KU...

2012-10-22T23:59:59.000Z

36

ESF Mine Power Center Platforms  

SciTech Connect (OSTI)

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.

T.A. Misiak

2000-02-10T23:59:59.000Z

37

MONITORED GEOLOGIC REPOSITORY SYSTEMS REQUIREMENTS DOCUMENT  

SciTech Connect (OSTI)

This document establishes the Monitored Geologic Repository system requirements for the U.S. Department of Energy's (DOE's) Civilian Radioactive Waste Management System (CRWMS). These requirements are based on the ''Civilian Radioactive Waste Management System Requirements Document'' (CRD) (DOE 2004a). The ''Monitored Geologic Repository Systems Requirements Document'' (MGR-RD) is developed in accordance with LP-3.3 SQ-OCRWM, ''Preparation, Review, and Approval of Office of Repository Development Requirements Document''. As illustrated in Figure 1, the MGR-RD forms part of the DOE Office of Civilian Radioactive Waste Management Technical Requirements Baseline. Revision 0 of this document identifies requirements for the current phase of repository design that is focused on developing a preliminary design for the repository and will be included in the license application submitted to the U.S. Nuclear Regulatory Commission for a repository at Yucca Mountain in support of receiving a construction authorization and subsequent operating license. As additional information becomes available, more detailed requirements will be identified in subsequent revisions to this document.

V. Trebules

2006-06-02T23:59:59.000Z

38

Modeling for Airborne Contamination  

SciTech Connect (OSTI)

The objective of Modeling for Airborne Contamination (referred to from now on as ''this report'') is to provide a documented methodology, along with supporting information, for estimating the release, transport, and assessment of dose to workers from airborne radioactive contaminants within the Monitored Geologic Repository (MGR) subsurface during the pre-closure period. Specifically, this report provides engineers and scientists with methodologies for estimating how concentrations of contaminants might be distributed in the air and on the drift surfaces if released from waste packages inside the repository. This report also provides dose conversion factors for inhalation, air submersion, and ground exposure pathways used to derive doses to potentially exposed subsurface workers. The scope of this report is limited to radiological contaminants (particulate, volatile and gaseous) resulting from waste package leaks (if any) and surface contamination and their transport processes. Neutron activation of air, dust in the air and the rock walls of the drift during the preclosure time is not considered within the scope of this report. Any neutrons causing such activation are not themselves considered to be ''contaminants'' released from the waste package. This report: (1) Documents mathematical models and model parameters for evaluating airborne contaminant transport within the MGR subsurface; and (2) Provides tables of dose conversion factors for inhalation, air submersion, and ground exposure pathways for important radionuclides. The dose conversion factors for air submersion and ground exposure pathways are further limited to drift diameters of 7.62 m and 5.5 m, corresponding to the main and emplacement drifts, respectively. If the final repository design significantly deviates from these drift dimensions, the results in this report may require revision. The dose conversion factors are further derived by using concrete of sufficient thickness to simulate the drift walls. The gamma-ray scattering properties of concrete are sufficiently similar to those of the host rock and proposed insert material; use of concrete will have no significant impact on the conclusions. The information in this report is presented primarily for use in performing pre-closure radiological safety evaluations of radiological contaminants, but it may also be used to develop strategies for contaminant leak detection and monitoring in the MGR. Included in this report are the methods for determining the source terms and release fractions, and mathematical models and model parameters for contaminant transport and distribution within the repository. Various particle behavior mechanisms that affect the transport of contaminant are included. These particle behavior mechanisms include diffusion, settling, resuspension, agglomeration and other deposition mechanisms.

F.R. Faillace; Y. Yuan

2000-08-31T23:59:59.000Z

39

CARRIER/CASK HANDLING SYSTEM DESCRIPTION DOCUMENT  

SciTech Connect (OSTI)

The Carrier/Cask Handling System receives casks on railcars and legal-weight trucks (LWTs) (transporters) that transport loaded casks and empty overpacks to the Monitored Geologic Repository (MGR) from the Carrier/Cask Transport System. Casks that come to the MGR on heavy-haul trucks (HHTs) are transferred onto railcars before being brought into the Carrier/Cask Handling System. The system is the interfacing system between the railcars and LWTs and the Assembly Transfer System (ATS) and Canister Transfer System (CTS). The Carrier/Cask Handling System removes loaded casks from the cask transporters and transfers the casks to a transfer cart for either the ATS or CTS, as appropriate, based on cask contents. The Carrier/Cask Handling System receives the returned empty casks from the ATS and CTS and mounts the casks back onto the transporters for reshipment. If necessary, the Carrier/Cask Handling System can also mount loaded casks back onto the transporters and remove empty casks from the transporters. The Carrier/Cask Handling System receives overpacks from the ATS loaded with canisters that have been cut open and emptied and mounts the overpacks back onto the transporters for disposal. If necessary, the Carrier/Cask Handling System can also mount empty overpacks back onto the transporters and remove loaded overpacks from them. The Carrier/Cask Handling System is located within the Carrier Bay of the Waste Handling Building System. The system consists of cranes, hoists, manipulators, and supporting equipment. The Carrier/Cask Handling System is designed with the tooling and fixtures necessary for handling a variety of casks. The Carrier/Cask Handling System performance and reliability are sufficient to support the shipping and emplacement schedules for the MGR. The Carrier/Cask Handling System interfaces with the Carrier/Cask Transport System, ATS, and CTS as noted above. The Carrier/Cask Handling System interfaces with the Waste Handling Building System for building structures and space allocations. The Carrier/Cask Handling System interfaces with the Waste Handling Building Electrical System for electrical power.

E.F. Loros

2000-06-23T23:59:59.000Z

40

Experimental Test Plan for Grouting H-3 Calcine  

SciTech Connect (OSTI)

Approximately 4400 cubic meters of solid high-level waste called calcine are stored at the Idaho Nuclear Technology and Engineering Center. Under the Idaho Cleanup Project, dual disposal paths are being investigated. The first path includes calcine retrieval, package "as-is", and ship to the Monitored Geological Repository (MGR). The second path involves treatment of the calcine with such methods as vitrification or grouting. This test plan outlines the hot bench scale tests to grout actual calcine and verify that the waste form properties meet the waste acceptance criteria. This is a necessary sequential step in the process of qualifying a new waste form for repository acceptance. The archive H-3 calcine samples at the Contaminated Equipment Maintenance Building attached to New Waste Calcining Facility will be used in these tests at the Remote Analytical Laboratory. The tests are scheduled for the second quarter of fiscal year 2007.

Alan K. Herbst

2006-01-01T23:59:59.000Z

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41

Enrico Fermi Fast Reactor Spent Nuclear Fuel Criticality Calculations: Degraded Mode  

SciTech Connect (OSTI)

The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the Department of Energy's (DOE) Enrico Fermi (EF) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP) and 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 the degraded mode internal configurations of the codisposal WP. The results of this calculation and those of Ref. 8 will be used to evaluate criticality issues and support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository.

D.R. Moscalu; L. Angers; J. Monroe-Rammsey; H.R. Radulesca

2000-07-21T23:59:59.000Z

42

Significant Radionuclides Determination  

SciTech Connect (OSTI)

The purpose of this calculation is to identify radionuclides that are significant to offsite doses from potential preclosure events for spent nuclear fuel (SNF) and high-level radioactive waste expected to be received at the potential Monitored Geologic Repository (MGR). In this calculation, high-level radioactive waste is included in references to DOE SNF. A previous document, ''DOE SNF DBE Offsite Dose Calculations'' (CRWMS M&O 1999b), calculated the source terms and offsite doses for Department of Energy (DOE) and Naval SNF for use in design basis event analyses. This calculation reproduces only DOE SNF work (i.e., no naval SNF work is included in this calculation) created in ''DOE SNF DBE Offsite Dose Calculations'' and expands the calculation to include DOE SNF expected to produce a high dose consequence (even though the quantity of the SNF is expected to be small) and SNF owned by commercial nuclear power producers. The calculation does not address any specific off-normal/DBE event scenarios for receiving, handling, or packaging of SNF. The results of this calculation are developed for comparative analysis to establish the important radionuclides and do not represent the final source terms to be used for license application. This calculation will be used as input to preclosure safety analyses and is performed in accordance with procedure AP-3.12Q, ''Calculations'', and is subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (DOE 2000) as determined by the activity evaluation contained in ''Technical Work Plan for: Preclosure Safety Analysis, TWP-MGR-SE-000010'' (CRWMS M&O 2000b) in accordance with procedure AP-2.21Q, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities''.

Jo A. Ziegler

2001-07-31T23:59:59.000Z

43

BWR Source Term Generation and Evaluation  

SciTech Connect (OSTI)

This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-00006 REV 01. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operating conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 7.28). The performance of the calculation and development of this document are carried out in accordance with AP-3.124, ''Design Calculation and Analyses'' (Ref. 7.29).

J.C. Ryman

2003-07-31T23:59:59.000Z

44

Identification of Aircraft Hazards  

SciTech Connect (OSTI)

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).

K. Ashley

2006-12-08T23:59:59.000Z

45

Evaluation of Blade-Strike Models for Estimating the Biological Performance of Kaplan Turbines  

SciTech Connect (OSTI)

Bio-indexing of hydroturbines is an important means to optimize passage conditions for fish by identifying operations for existing and new design turbines that minimize the probability of injury. Cost-effective implementation of bio-indexing requires the use of tools such as numerical and physical turbine models to generate hypotheses for turbine operations that can be tested at prototype scales using live fish. Numerical deterministic and stochastic blade strike models were developed for a 1:25-scale physical turbine model built by the U.S. Army Corps of Engineers for the original design turbine at McNary Dam and for prototype-scale original design and replacement minimum gap runner (MGR) turbines at Bonneville Dam's first powerhouse. Blade strike probabilities predicted by both models were comparable with the overall trends in blade strike probability observed in both prototype-scale live fish survival studies and physical turbine model using neutrally buoyant beads. The predictions from the stochastic model were closer to the experimental data than the predictions from the deterministic model because the stochastic model included more realistic consideration of the aspect of fish approaching to the leading edges of turbine runner blades. Therefore, the stochastic model should be the preferred method for the prediction of blade strike and injury probability for juvenile salmon and steelhead using numerical blade-strike models.

Deng, Zhiqun; Carlson, Thomas J.; Ploskey, Gene R.; Richmond, Marshall C.; Dauble, Dennis D.

2007-11-10T23:59:59.000Z

46

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

SciTech Connect (OSTI)

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).

D.R. Moscalu

2002-08-28T23:59:59.000Z

47

Preliminary Transportation, Aging and Disposal Canister System Performance Specification  

SciTech Connect (OSTI)

This document provides specifications for selected system components of the Transportation, Aging and Disposal (TAD) canister-based system. A list of system specified components and ancillary components are included in Section 1.2. The TAD canister, in conjunction with specialized overpacks will accomplish a number of functions in the management and disposal of spent nuclear fuel. Some of these functions will be accomplished at purchaser sites where commercial spent nuclear fuel (CSNF) is stored, and some will be performed within the Office of Civilian Radioactive Waste Management (OCRWM) transportation and disposal system. This document contains only those requirements unique to applications within Department of Energy's (DOE's) system. DOE recognizes that TAD canisters may have to perform similar functions at purchaser sites. Requirements to meet reactor functions, such as on-site dry storage, handling, and loading for transportation, are expected to be similar to commercially available canister-based systems. This document is intended to be referenced in the license application for the Monitored Geologic Repository (MGR). As such, the requirements cited herein are needed for TAD system use in OCRWM's disposal system. This document contains specifications for the TAD canister, transportation overpack and aging overpack. The remaining components and equipment that are unique to the OCRWM system or for similar purchaser applications will be supplied by others.

C.A Kouts

2006-11-22T23:59:59.000Z

48

DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION  

SciTech Connect (OSTI)

The purpose of this analysis/model report (AMR) is to select and/or abstract conservative degradation models for DOE-(US. Department of Energy) owned spent nuclear fuel (DSNF) and the immobilized ceramic plutonium (Pu) disposition waste forms for application in the proposed monitored geologic repository (MGR) postclosure Total System Performance Assessment (TSPA). Application of the degradation models abstracted herein for purposes other than TSPA should take into consideration the fact that they are, in general, very conservative. Using these models, the forward reaction rate for the mobilization of radionuclides, as solutes or colloids, away from the waste fondwater interface by contact with repository groundwater can then be calculated. This forward reaction rate generally consists of the dissolution reaction at the surface of spent nuclear fuel (SNF) in contact with water, but the degradation models, in some cases, may also include and account for the physical disintegration of the SNF matrix. The models do not, however, account for retardation, precipitation, or inhibition of the migration of the mobilized radionuclides in the engineered barrier system (EBS). These models are based on the assumption that all components of the DSNF waste form are released congruently with the degradation of the matrix.

T.A. Thornton

2000-12-20T23:59:59.000Z

49

WASTE PACKAGE OPERATIONS FY99 CLOSURE METHODS REPORT  

SciTech Connect (OSTI)

The waste package (WP) closure weld development task is part of a larger engineering development program to develop waste package designs. The purpose of the larger waste package engineering development program is to develop nuclear waste package fabrication and closure methods that the Nuclear Regulatory Commission will find acceptable and will license for disposal of spent nuclear fuel (SNF), non-fuel components, and vitrified high-level waste within a Monitored Geologic Repository (MGR). Within the WP closure development program are several major development tasks, which, in turn, are divided into subtasks. The major tasks include: WP fabrication development, WP closure weld development, nondestructive examination (NDE) development, and remote in-service inspection development. The purpose of this report is to present the objectives, technical information, and work scope relating to the WP closure weld development.and NDE tasks and subtasks and to report results of the closure weld and NDE development programs for fiscal year 1999 (FY-99). The objective of the FY-99 WP closure weld development task was to develop requirements for closure weld surface and volumetric NDE performance demonstrations, investigate alternative NDE inspection techniques, and develop specifications for welding, NDE, and handling system integration. In addition, objectives included fabricating several flat plate mock-ups that could be used for NDE development, stress relief peening, corrosion testing, and residual stress testing.

M. C. Knapp

1999-09-23T23:59:59.000Z

50

REVIEW OF NRC APPROVED DIGITAL CONTROL SYSTEMS ANALYSIS  

SciTech Connect (OSTI)

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.

D.W. Markman

1999-09-17T23:59:59.000Z

51

Initial Cladding Condition  

SciTech Connect (OSTI)

The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M&O 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in evaluating the post-closure performance of the Monitored Geologic Repository (MGR) in relation to waste form degradation.

E. Siegmann

2000-08-22T23:59:59.000Z

52

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages  

SciTech Connect (OSTI)

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to time periods up to 6.35 x 10{sup 5} years. This longer time frame is closer to the one million year time horizon recently recommended by the National Academy of Sciences to the Environmental Protection Agency for performance assessment related to a nuclear repository (Ref. 5). However, it is important to note that after 100,000 years, most of the materials of interest (fissile materials) will have either been removed from the WP, reached a steady state, or been transmuted.

P. Bernot

2001-02-27T23:59:59.000Z

53

Subsurface Contamination Control  

SciTech Connect (OSTI)

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.

Y. Yuan

2001-11-16T23:59:59.000Z

54

Subsurface Contamination Control  

SciTech Connect (OSTI)

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.

Y. Yuan

2001-12-12T23:59:59.000Z

55

WHB/WTB SPACE PROGRAM ANALYSIS FOR SITE RECOMMENDATION  

SciTech Connect (OSTI)

The purpose of this analysis is to identify and evaluate the functional space and spatial relationship requirements for the two main nuclear buildings, the Waste Handling Building (WHB) and the Waste Treatment Building (WTB), which are part of the Repository Surface Facilities. This analysis is consistent with the Development Plan for ''WHB/WTB Space Program Analysis for Site Recommendation'' (CRWMS M&O 2000r), which concentrates on the primary, primary support, facility support, and miscellaneous building support areas located in the WHB and WTB. The development plan was completed in accordance with AP-2.134, ''Technical Product Development Planning''. The objective and scope of this analysis is to develop a set of spatial parameters (e.g., square footage, room heights, etc.) and layout requirements (e.g., adjacency and access/circulation requirements, etc.) from which preliminary building floor plans are developed and presented as figures. The resulting figures will provide information to support the Site Recommendation and the total system life cycle cost. This analysis uses the Viability Assessment (VA) ''Surface Nuclear Facilities Space Program Analysis'' (SPA) (CRWMS M&O 1997c) as the baseline reference document and further develops the functional requirements based on Project-directed changes, including incorporation of a new design basis waste stream and the applicable elements of Enhanced Design Alternative (EDA)-II, as identified in the ''License Application Design Selection Report'' (CRWMS M&O 1999e), which followed the initial SPA (baseline). The impacts of the EDA-II were almost entirely to the WHB. To meet the EDA-II thermal requirements, hotter fuel would be handled, therefore requiring a fuel-blending pool to be added to the WHB in order to age the hotter he1 at the repository and provide for commercial spent nuclear fuel (CSNF) blending. In addition to EDA-II recommendations, the waste stream was modified, including the elimination of approximately 300 multi-purpose canisters from the CSNF schedule. The bases for the Monitored Geologic Repository (MGR) surface design changes, as a result of the waste stream changes, are defined in ''Calculations from Surface Facilities Operations in Support of the Revision to the Waste Quantity, Mix, and Throughput Study'' (CRWMS M&O 2000c, Section 2.4). This effort resulted in a reduction in the number of canister transfer lines from 2 to 1. In addition, as indicated in ''WITNESS Model Input for Thermal Blending of Commercial Spent Nuclear Fuel Assemblies'' (CRWMS M&O 19991), the quantity of dual-purpose canisters (DPCs) assumed to be shipped to the repository has been reduced by about 37 percent. This change resulted in a reduction of the number of Assembly Transfer System (ATS) lines in the WHB from 3 to 2. In summary, this analysis is intended to provide a preliminary level of design showing room square footages and heights associated with the WHB and WTB. These spatial dimensions are anticipated to increase or decrease as the design progresses.

W.D. Lindholm

2000-05-25T23:59:59.000Z

56

DOE Hydropower Program Biennial Report for FY 2005-2006  

SciTech Connect (OSTI)

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.

Sale, Michael J [ORNL; Cada, Glenn F [ORNL; Acker, Thomas L. [Northern Arizona State University and National Renewable Energy Laboratory; Carlson, Thomas [Pacific Northwest National Laboratory (PNNL); Dauble, Dennis D. [Pacific Northwest National Laboratory (PNNL); Hall, Douglas G. [Idaho National Laboratory (INL)

2006-07-01T23:59:59.000Z

57

Uncanistered Spent Nuclear fuel Disposal Container System Description Document  

SciTech Connect (OSTI)

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.

NONE

2000-10-12T23:59:59.000Z

58

EQ6 Calculation for Chemical Degradation of Shippingport LWBR (TH/U Oxide) Spent Nuclear Fuel Waste Packages  

SciTech Connect (OSTI)

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site. Because of the high content of fissile material in the SNF, the waste package (WP) design requires special consideration of the amount and placement of neutron absorbers and the possible loss of absorbers and SNF materials over geologic time. For some WPs, the outer shell corrosion-resistant material (CRM) and the corrosion-allowance inner shell may breach (Refs. 2 and 3), allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components and neutron absorbers from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing a Shippingport LWBR SNF seed assembly, and high-level waste (HLW) glass canisters arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which criticality control material, suggested for this WP design, will remain in the WP after corrosion/dissolution of the initial WP configuration (such that it can be effective in preventing criticality); (2) The extent to which fissile uranium and fertile thorium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations), of the simulations are limited to time periods up to 3.17 x 10{sup 5} years. This longer time frame is closer to the one million year time horizon recently recommended by the National Academy of Sciences to the Environmental Protection Agency for performance assessment related to a nuclear repository (Ref. 5). However, it is important to note that after 100,000 years, most of the materials of interest (fissile and absorber materials) will have either been removed from the WP, reached a steady state, or been transmuted. The calculation included elements with high neutron-absorption cross sections, notably gadolinium (Gd), as well as the fissile materials. The results of this analysis will be used to ensure that the type and amount of criticality control material used in the WP design will prevent criticality.

S. Arthur

2000-09-14T23:59:59.000Z