'CANDLE' burnup regime after LWR regime

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Nagata, Akito

2008-01-01

CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)

Civilian Radioactive Waste Management System Requirements Document

International Nuclear Information System (INIS)

1992-12-01

This document specifies the top-level requirements for the Civilian Radioactive Waste Management System (CRWMS). The document is referred to herein as the CRD, for CRWMS Requirements document. The OCRWM System Engineering Management Plan (SEMP) establishes the technical document hierarchy (hierarchy of technical requirements and configuration baseline documents) for the CRWMS program. The CRD is the top-level document in this hierarchy. The immediate subordinate documents are the System Requirements Documents (SRDS) for the four elements of the CRWMS and the Interface Specification (IFS). The four elements of the CRWMS are the Waste Acceptance System, the Transportation System, the Monitored Retrievable Storage (MRS) System and the Mined Geologic Disposal System (MGDS). The Interface Specification describes the six inter-element interfaces between the four elements. This hierarchy establishes the requirements to be addressed by the design of the system elements. Many of the technical requirements for the CRWMS are documented in a variety of Federal regulations, DOE directives and other Government documentation. It is the purpose of the CRD to establish the technical requirements for the entire program. In doing so, the CRD summarizes source documentation for requirements that must be addressed by the program, specifies particular requirements, and documents derived requirements that are not covered in regulatory and other Government documentation, but are necessary to accomplish the mission of the CRWMS. The CRD defines the CRWMS by identifying the top-level functions the elements must perform (These top-level functions were derived using functional analysis initially documented in the Physical System Requirements (PSR) documents). The CRD also defines the top-level physical architecture of the system and allocates the functions and requirements to the architectural elements of the system

Mined Geologic Disposal System Requirements Document

International Nuclear Information System (INIS)

1993-01-01

This Mined Geologic Disposal System Requirements document (MGDS-RD) describes the functions to be performed by, and the requirements for, a Mined Geologic Disposal System (MGDS) for the permanent disposal of spent nuclear fuel (SNF) and commercial and defense high level radioactive waste (HLW) in support of the Civilian Radioactive Waste Management System (CRWMS). The development and control of the MGDS-RD is quality-affecting work and is subject to the Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) Quality Assurance Requirements Document (QARD). As part of the technical requirements baseline, it is also subject to Baseline Management Plan controls. The MGDS-RD and the other program-level requirements documents have been prepared and managed in accordance with the Technical Document Preparation Plan (TDPP) for the Preparation of System Requirements Documents

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

International Nuclear Information System (INIS)

Greene, S.R.; Fisher, S.E.; Bevard, B.B.

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

Energy Technology Data Exchange (ETDEWEB)

Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

Implications of plutonium utilization strategies on the transition from a LWR economy to a breeder economy

International Nuclear Information System (INIS)

Newman, D.F.; Fleischman, R.M.; White, M.K.

1977-02-01

The plutonium interface between the LWR and LMFBR fuel cycles is examined for typical nuclear growth projections both with and without plutonium recycle in LWRs. In order to guarantee a fuel supply for projected LMFBR growth rates, significant multiple Pu recycle in LWRs will not be possible. However, about 78% of the benefit of multiple plutonium recycle between now and the turn of the century is realized by one recycle and then stockpiling spent MOX for the LMFBR. LMFBR reprocessing schecules are estimated based on accumulation of reprocessing load. These schedules are used to estimate the amount of plutonium recovered from LMFBR fuels and determine the residual LWR plutonium required to meet LMFBR demand. The stockpile of LWR produced plutonium in spent MOX is sufficient to fuel the LMFBR until commercial LMFBR reprocessing can be justified. After that time, recycle of plutonium in LWRs will be significantly limited by a continuing LMFBR demand for LWR plutonium due to the projected high LMFBR growth rate. LWR reprocessing requirements are estimated for the assumed condition that LWR plutonium recycle is not approved, but the LMFBR is still pursued as an energy option. The uncertainties presented by this condition are addressed qualitatively. However, in our judgment these uncertainties in the plutonium market would likely delay LMFBR growth to levels significantly below current projections

Waste Management System Requirement document

International Nuclear Information System (INIS)

1990-04-01

This volume defines the top level technical requirements for the Monitored Retrievable Storage (MRS) facility. It is designed to be used in conjunction with Volume 1, General System Requirements. Volume 3 provides a functional description expanding the requirements allocated to the MRS facility in Volume 1 and, when appropriate, elaborates on requirements by providing associated performance criteria. Volumes 1 and 3 together convey a minimum set of requirements that must be satisfied by the final MRS facility design without unduly constraining individual design efforts. The requirements are derived from the Nuclear Waste Policy Act of 1982 (NWPA), the Nuclear Waste Policy Amendments Act of 1987 (NWPAA), the Environmental Protection Agency's (EPA) Environmental Standards for the Management and Disposal of Spent Nuclear Fuel (40 CFR 191), NRC Licensing Requirements for the Independent Storage of Spent Nuclear and High-Level Radioactive Waste (10 CFR 72), and other federal statutory and regulatory requirements, and major program policy decisions. This document sets forth specific requirements that will be fulfilled. Each subsequent level of the technical document hierarchy will be significantly more detailed and provide further guidance and definition as to how each of these requirements will be implemented in the design. Requirements appearing in Volume 3 are traceable into the MRS Design Requirements Document. Section 2 of this volume provides a functional breakdown for the MRS facility. 1 tab

TWRS configuration management requirement source document

International Nuclear Information System (INIS)

Vann, J.M.

1997-01-01

The TWRS Configuration Management (CM) Requirement Source document prescribes CM as a basic product life-cycle function by which work and activities are conducted or accomplished. This document serves as the requirements basis for the TWRS CM program. The objective of the TWRS CM program is to establish consistency among requirements, physical/functional configuration, information, and documentation for TWRS and TWRS products, and to maintain this consistency throughout the life-cycle of TWRS and the product, particularly as changes are being made

Mined Geologic Disposal System Requirements Document

International Nuclear Information System (INIS)

1994-03-01

This Mined Geologic Disposal System Requirements Document (MGDS-RD) describes the functions to be performed by, and the requirements for, a Mined Geologic Disposal System (MGDS) for the permanent disposal of spent nuclear fuel (SNF) (including SNF loaded in multi-purpose canisters (MPCs)) and commercial and defense high-level radioactive waste (HLW) in support of the Civilian Radioactive Waste Management System (CRWMS). The purpose of the MGDS-RD is to define the program-level requirements for the design of the Repository, the Exploratory Studies Facility (ESF), and Surface Based Testing Facilities (SBTF). These requirements include design, operation, and decommissioning requirements to the extent they impact on the physical development of the MGDS. The document also presents an overall description of the MGDS, its functions (derived using the functional analysis documented by the Physical System Requirements (PSR) documents as a starting point), its segments as described in Section 3.1.3, and the requirements allocated to the segments. In addition, the program-level interfaces of the MGDS are identified. As such, the MGDS-RD provides the technical baseline for the design of the MGDS

LWR-core behaviour project

International Nuclear Information System (INIS)

Paratte, J.M.

1982-07-01

The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis

International Nuclear Information System (INIS)

Ivanov, Kostadin; Avramova, Maria

2007-01-01

The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical

Recycling U and Pu in LWR

International Nuclear Information System (INIS)

Zheng Hualing.

1986-01-01

This article, from viewpoints of technical feasibility, safety evaluation and socioeconomic benefit-risk analysis, introduces and comments on history and status of recycling U and Pu in LWR, dealing with reactor, reprocessing, conversion and fuel element fabrication et al. Author has analysed LWR fuel cycle strategies in China and made a proposal

Waste Acceptance System Requirements document (WASRD)

International Nuclear Information System (INIS)

1993-01-01

This Waste Acceptance System Requirements document (WA-SRD) describes the functions to be performed and the technical requirements for a Waste Acceptance System for accepting spent nuclear fuel (SNF) and high-level radioactive waste (HLW) into the Civilian Radioactive Waste Management System (CRWMS). This revision of the WA-SRD addresses the requirements for the acceptance of HLW. This revision has been developed as a top priority document to permit DOE's Office of Environmental Restoration and Waste Management (EM) to commence waste qualification runs at the Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) in a timely manner. Additionally, this revision of the WA-SRD includes the requirements from the Physical System Requirements -- Accept Waste document for the acceptance of SNF. A subsequent revision will fully address requirements relative to the acceptance of SNF

EURLIB-LWR-45/16 and - 15/5. Two board group libraries for LWR-shielding problems

Energy Technology Data Exchange (ETDEWEB)

Herrnberger, V

1982-04-01

Specifications of the broad group cross section libraries EURLIB-LWR-45/16 and -15/5 are given. They are based on EURLIB-III data and produced for LWR shielding problems. The elements considered are H, C{sub 12}, O, Na, Al, Si, Ca, Cr, Mn, Fe, Ni, Zr, U{sub 235}, U{sub 238}. The cross section libraries are available upon request from EIR, RSIC, NEA-CPL and IAEA-NDS. (author) Refs, figs, tabs

European utility requirements: common rules to design next LWR plants in an open electricity market

International Nuclear Information System (INIS)

Berbey, Pierre; Ingemarsson, Karl-Fredrik

2004-01-01

The major European electricity producers want to keep able to build new nuclear power plants and they believe 3. generation LWRs would be the most adapted response to their needs in the first decades of this century. Producing a common European Utility Requirement (EUR) document has been one of the basic tasks towards this objective. In this common frame, standardized and competitive LWR NPPs could be developed and offered to the investors. This idea is now well supported by all the other actors on the European electricity market: vendors, regulators, grid managers, administrations although in the competitive and unified European electricity market that is emerging, the electricity producers' stakes are more and more different from the other electricity business actors'. The next term objectives of the electricity producers involved in EUR are focused on negotiating common rules of the game together with the regulators. This covers the nuclear safety approaches, the conditions requested to connect a plant to a HV grid, as well as the design standards. Discussions are going on between the EUR organization and all the corresponding bodies to develop stabilized and predictable design rules that would meet the constraints of nuclear electricity generation in this new environment. Finally there cannot be competition without competitors. The EUR organization has proven to be the right place to establish trustful relationship between the vendors and their potential customers, through fair assessment of the proposed designs performance vs. the utility needs. This will be continued and developed with the main vendors present in Europe, so as to keep alive a list of 4 to 6 designs 'qualified', i.e. showing an acceptable score of non-compliance vs. EUR. (authors)

Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2B, User's guide to the LWR assemblies data base, Appendix 2C, User's guide to the LWR radiological data base, Appendix 2D, User's guide to the LWR quantities data base

International Nuclear Information System (INIS)

1987-12-01

This User's Guide for the LWR Assemblies data base system is part of the Characteristics Data Base being developed under the Waste Systems Data Development Program. The objective of the LWR Assemblies data base is to provide access at the personal computer level to information about fuel assemblies used in light-water reactors. The information available is physical descriptions of intact fuel assemblies and radiological descriptions of spent fuel disassembly hardware. The LWR Assemblies data base is a user-oriented menu driven system. Each menu is instructive about its use. Section 5 of this guide provides a sample session with the data base to assist the user

Tank waste remediation system functions and requirements document

Energy Technology Data Exchange (ETDEWEB)

Carpenter, K.E

1996-10-03

This is the Tank Waste Remediation System (TWRS) Functions and Requirements Document derived from the TWRS Technical Baseline. The document consists of several text sections that provide the purpose, scope, background information, and an explanation of how this document assists the application of Systems Engineering to the TWRS. The primary functions identified in the TWRS Functions and Requirements Document are identified in Figure 4.1 (Section 4.0) Currently, this document is part of the overall effort to develop the TWRS Functional Requirements Baseline, and contains the functions and requirements needed to properly define the top three TWRS function levels. TWRS Technical Baseline information (RDD-100 database) included in the appendices of the attached document contain the TWRS functions, requirements, and architecture necessary to define the TWRS Functional Requirements Baseline. Document organization and user directions are provided in the introductory text. This document will continue to be modified during the TWRS life-cycle.

Tank waste remediation system functions and requirements document

International Nuclear Information System (INIS)

Carpenter, K.E

1996-01-01

This is the Tank Waste Remediation System (TWRS) Functions and Requirements Document derived from the TWRS Technical Baseline. The document consists of several text sections that provide the purpose, scope, background information, and an explanation of how this document assists the application of Systems Engineering to the TWRS. The primary functions identified in the TWRS Functions and Requirements Document are identified in Figure 4.1 (Section 4.0) Currently, this document is part of the overall effort to develop the TWRS Functional Requirements Baseline, and contains the functions and requirements needed to properly define the top three TWRS function levels. TWRS Technical Baseline information (RDD-100 database) included in the appendices of the attached document contain the TWRS functions, requirements, and architecture necessary to define the TWRS Functional Requirements Baseline. Document organization and user directions are provided in the introductory text. This document will continue to be modified during the TWRS life-cycle

HFR irradiation testing of light water reactor (LWR) fuel

International Nuclear Information System (INIS)

Markgraf, J.F.W.

1985-01-01

For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

Nondestructive evaluation of LWR spent fuel shipping casks

International Nuclear Information System (INIS)

Ballard, D.W.

1978-02-01

An analysis of nondestructive testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined

Monitored Retrievable Storage System Requirements Document

International Nuclear Information System (INIS)

1994-03-01

This Monitored Retrievable Storage System Requirements Document (MRS-SRD) describes the functions to be performed and technical requirements for a Monitored Retrievable Storage (MRS) facility subelement and the On-Site Transfer and Storage (OSTS) subelement. The MRS facility subelement provides for temporary storage, at a Civilian Radioactive Waste Management System (CRWMS) operated site, of spent nuclear fuel (SNF) contained in an NRC-approved Multi-Purpose Canister (MPC) storage mode, or other NRC-approved storage modes. The OSTS subelement provides for transfer and storage, at Purchaser sites, of spent nuclear fuel (SNF) contained in MPCs. Both the MRS facility subelement and the OSTS subelement are in support of the CRWMS. The purpose of the MRS-SRD is to define the top-level requirements for the development of the MRS facility and the OSTS. These requirements include design, operation, and decommissioning requirements to the extent they impact on the physical development of the MRS facility and the OSTS. The document also presents an overall description of the MRS facility and the OSTS, their functions (derived by extending the functional analysis documented by the Physical System Requirements (PSR) Store Waste Document), their segments, and the requirements allocated to the segments. In addition, the top-level interface requirements of the MRS facility and the OSTS are included. As such, the MRS-SRD provides the technical baseline for the MRS Safety Analysis Report (SAR) design and the OSTS Safety Analysis Report design

Evaluation of nuclear fuel reprocessing strategies. 2. LWR fuel storage, recycle economics and plutonium logistics

International Nuclear Information System (INIS)

Prince, B.E.; Hadley, S.W.

1983-01-01

This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of the time-dependent aspects of plutonium requirements and supply relative to this transition

22 CFR 40.71 - Documentation requirements for immigrants.

Science.gov (United States)

2010-04-01

... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Documentation requirements for immigrants. 40... NONIMMIGRANTS AND IMMIGRANTS UNDER THE IMMIGRATION AND NATIONALITY ACT, AS AMENDED Documentation Requirements § 40.71 Documentation requirements for immigrants. INA 212(a)(7)(A) is not applicable at the time of...

Is it the end of history for LWR safety?

International Nuclear Information System (INIS)

Sehgal, Bal Raj

2004-01-01

In this essay a parallel is drawn between the struggle for recognition, which is argued by Fukuyama as the 'motor' of human history and that waged by the LWR safety for the public to recognize the LWR plants as a source of safe nuclear power. The end of history for the ''human struggle for recognition'' as the capitalistic liberal democracy is equated with the ''end of history'' for the LWR safety to provide assurance to the public of termination of a severe accident it ever would occur. It is suggested that we are near ''the end of history'' of the LWR safety for the new-design LWR plants but fall short for the presently-installed plants. The essay bases these suggestions on an examination of the history of nuclear power development in U.S.A., but also considering the more recent regulatory and public acceptance developments in Europe and the rest of the World. (author)

Perspectives on the economic risks of LWR accidents

International Nuclear Information System (INIS)

Ritchie, L.T.; Burke, R.P.

1986-01-01

Models which can be used for the analysis of the economic risks from events which may occur during LWR operation have been developed. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The economic consequence models have been applied in studies of the economic risks from the operation of US LWR plants. The results of the analyses provide some important perspectives regarding the economic risks of LWR accidents. The analyses indicate that economic risks, in contrast to public health risks, are dominated by the onsite costs of relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for a typical US plant

Implementation of static generalized perturbation theory for LWR design applications

International Nuclear Information System (INIS)

Byron, R.F.; White, J.R.

1987-01-01

A generalized perturbation theory (GPT) formulation is developed for application to light water reactor (LWR) design. The extensions made to standard generalized perturbation theory are the treatment of thermal-hydraulic and fission product poisoning feedbacks, and criticality reset. This formulation has been implemented into a standard LWR design code. The method is verified by comparing direct calculations with GPT calculations. Data are presented showing that feedback effects need to be considered when using GPT for LWR problems. Some specific potential applications of this theory to the field of LWR design are discussed

FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

Recycle of LWR actinides to an IFR

International Nuclear Information System (INIS)

Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

1991-01-01

Large quantities of actinide elements are present in irradiated light water reactor fuel that is stored throughout the world. Because of the high fission to capture ratio for the transuranium (TRU) elements with the high energy neutrons in metal-fueled integral fast reactors (IFR), that reactor can consume these elements effectively. The stored fuel may represent valuable resource for the expanding application of fast power reactors. In addition, the removal of TRU elements from spent LWR fuel has the potential for increasing the capacity of high level waste facilities by reducing the heat load and may increase the margin of safety in meeting licensing requirement. Argonne National Laboratory is developing a pyrochemical process, which is compatible with the IFR fuel cycle for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. Two pyrochemical processes, that is, salt transport process and blanket processing study, are discussed in this paper. Also the experimental studies are reported. (K.I.)

Technical requirements document for the waste flow analysis

International Nuclear Information System (INIS)

Shropshire, D.E.

1996-05-01

Purpose of this Technical Requirements Document is to define the top level customer requirements for the Waste Flow Analysis task. These requirements, once agreed upon with DOE, will be used to flow down subsequent development requirements to the model specifications. This document is intended to be a ''living document'' which will be modified over the course of the execution of this work element. Initial concurrence with the technical functional requirements from Environmental Management (EM)-50 is needed before the work plan can be developed

Fiscal year 1999 waste information requirements document

International Nuclear Information System (INIS)

Adams, M.R.

1998-01-01

The Waste Information Requirements Document (WIRD) has the following purposes: To describe the overall drivers that require characterization information and to document their source; To define how characterization is going to satisfy the drivers, close issues, and measure and report progress; and To describe deliverables and acceptance criteria for characterization. Characterization information is required to maintain regulatory compliance, perform operations and maintenance, resolve safety issues, and prepare for disposal of waste. Commitments addressing these requirements are derived from the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement; the Recommendation 93-5 Implementation Plan (DOE-RL 1996a) to the Defense Nuclear Facilities Safety Board (DNFSB); and other requirement sources listed in Section 2.0. The Waste Information Requirements Document replaces the tank waste analysis plans and the tank characterization plan previously required by the Tri-Party Agreement, Milestone M-44-01 and M-44-02 series

Creep damage in zircaloy-4 at LWR temperatures

International Nuclear Information System (INIS)

Keusseyan, R.L.; Hu, C.P.; Li, C.Y.

1978-08-01

The observation of creep damage in the form of grain boundary cavitation in Zircaloy-4 in the temperature range of interest to Light Water Reactor (LWR) applications is reported. The observed damage is shown to reduce the ductility of Zircaloy-4 in a tensile test at LWR temperatures

Fatigue management considering LWR coolant environments

International Nuclear Information System (INIS)

Park, Heung Bae; Jin, Tae eun

2000-01-01

Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

Flexible fuel cycle system for the transition from LWR to FBR

International Nuclear Information System (INIS)

Fukasawa, Tetsuo; Yamashita, Junichi; Hoshino, Kuniyoshi; Sasahira, Akira; Inoue, Tadashi; Minato, Kazuo; Sato, Seichi

2009-01-01

Japan will deploy commercial fast breeder reactor (FBR) from around 2050 under the suitable conditions for the replacement of light water reactor (LWR) with FBR. The transition scenario from LWR to FBR is investigated in detail and the flexible fuel cycle initiative (FFCI) system has been proposed as a optimum transition system. The FFCI removes ∼95% uranium from LWR spent fuel (SF) in LWR reprocessing and residual material named Recycle Material (RM), which is ∼1/10 volume of original SF and contains ∼50% U, ∼10% Pu and ∼40% other nuclides, is treated in FBR reprocessing to recover Pu and U. If the FBR deployment speed becomes lower, the RM will be stored until the higher speed again. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR SF during the transition period. The economy is better for FFCI due to the smaller LWR reprocessing facility (no Pu/U recovery and fabrication). The FFCI can supply high Pu concentration RM, which has high proliferation resistance and flexibly respond to FBR introduction rate changes. Volume minimization of LWR SF is possible for FFCI by its conversion to RM. Several features of FFCI were quantitatively evaluated such as Pu mass balance, reprocessing capacities, LWR SF amounts, RM amounts, and proliferation resistance to compare the effectiveness of the FFCI system with other systems. The calculated Pu balance revealed that the FFCI could supply enough but no excess Pu to FBR. These evaluations demonstrated the applicability of FFCI system to the transition period from LWR to FBR cycles. (author)

Documentation requirements for radiation sterilization

DEFF Research Database (Denmark)

Miller, A.

1995-01-01

Several standards are recently approved or are under development by the standard organizations ISO and CEN in the field of radiation sterilization. Particularly in Europe these standards define new requirements on some issues and on other issues they emphasize the necessary documentation for appr......Several standards are recently approved or are under development by the standard organizations ISO and CEN in the field of radiation sterilization. Particularly in Europe these standards define new requirements on some issues and on other issues they emphasize the necessary documentation...... for approval of radiation sterilized products. The impact of these standards on the radiation sterilization is discussed, with special attention given to a few special issues, mainly traceability and uncertainty of measurement results....

FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

International Nuclear Information System (INIS)

1996-09-01

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ''Alternative Teams,'' chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S ampersand S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT's analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option

Environmental Restoration Remedial Action quality assurance requirements document

International Nuclear Information System (INIS)

1991-01-01

This document defines the quality assurance requirements for the US Department of Energy-Richland Operations Office Environmental Restoration Remedial Action program at the Hanford Site. The Environmental Restoration Remedial Action program implements significant commitments made by the US Department of Energy in the Hanford Federal Facility Agreement and Consent Order entered into with the Washington State Department of Ecology and the US Environmental Protection Agency. This document combines quality assurance requirements from various source documents into one set of requirements for use by the US Department of Energy-Richland Operations Office and other Environmental Restoration Remedial Action program participants. This document will serve as the basis for developing Quality Assurance Program Plans and implementing procedures by the participants. The requirements of this document will be applied to activities affecting quality, using a graded approach based on the importance of the item, service, or activity to the program objectives. The Quality Assurance Program that will be established using this document as the basis, together with other program and technical documents, form an integrated management control system for conducting the Environmental Restoration Remedial Action program activities in a manner that provides safety and protects the environment and public health

Fluor Daniel Hanford contract standards/requirements identification document

Energy Technology Data Exchange (ETDEWEB)

Bennett, G.L.

1997-04-24

This document, the Standards/Requirements Identification Document (S/RID) for the Fluor Daniel Hanford Contract, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment.

Fluor Daniel Hanford company standards requirements identification document

International Nuclear Information System (INIS)

Bennett, G.L.

1997-01-01

This document, the Standards/Requirements Identification Document (S/RID) for the Fluor Daniel Hanford Contract, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment

Transportation system requirements document. Revision 1 DCN01. Supplement

International Nuclear Information System (INIS)

1995-05-01

The original Transportation System Requirements Document described the functions to be performed by and the technical requirements for the Transportation System to transport spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from Purchaser and Producer sites to a Civilian Radioactive Waste Management System (CRWMS) site, and between CRWMS sites. The purpose of that document was to define the system-level requirements. These requirements include design and operations requirements to the extent they impact on the development of the physical segments of Transportation. The document also presented an overall description of Transportation, its functions, its segments, and the requirements allocated to the segments and the system-level interfaces with Transportation. This revision of the document contains only the pages that have been modified

Materials choices for the advanced LWR steam generators

International Nuclear Information System (INIS)

Paine, J.P.N.; Shoemaker, C.E.; McIlree, A.R.

1987-01-01

Current light water reactor (LWR) steam generators have been affected by a variety of corrosion and mechanical damage degradation mechanisms. Included are wear caused by tube vibration, intergranular corrosion, pitting, and thinning or wastage of the steam generator tubing and accelerated corrosion of carbon steel supports (denting). The Electric Power Research Institute (EPRI) and the Steam Generator Owners Groups (I, II) have sponsored laboratory and field studies to provide ameliorative actions for the majority of the damage forms experienced to date. Some of the current corrosion mechanisms are aggravated or caused by unique materials choices or materials interactions. New materials have been proposed and at least partially qualified for use in replacement model steam generators, including an advanced LWR design. In so far as possible, the materials choices for the advanced LWR steam generator avoid the corrosion pitfalls seemingly inherent in the current designs. The EPRI Steam Generator Project staff has recommended materials and design choices for a new steam generator. Based on these recommendations we believe that the advanced LWR steam generators will be much less affected by corrosion and mechanical damage mechanisms than are now experienced

Modeling the economic consequences of LWR accidents

International Nuclear Information System (INIS)

Burke, R.P.; Aldrich, D.C.; Rasmussen, N.C.

1984-01-01

Models to be used for analyses of economic risks from events which may occur during LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models can be used by both the nuclear power industry and regulatory agencies in cost-benefit analyses for decisionmaking purposes. The newly developed economic consequence models are applied in an example to estimate the economic risks from operation of the Surry Unit 2 plant. The analyses indicate that economic risks from US LWR operation, in contrast to public health risks, are dominated by relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for the Surry site. The implications of these conclusions for nuclear power plant operation and regulation are discussed

Spectrum analysis on quality requirements consideration in software design documents.

Science.gov (United States)

Kaiya, Haruhiko; Umemura, Masahiro; Ogata, Shinpei; Kaijiri, Kenji

2013-12-01

Software quality requirements defined in the requirements analysis stage should be implemented in the final products, such as source codes and system deployment. To guarantee this meta-requirement, quality requirements should be considered in the intermediate stages, such as the design stage or the architectural definition stage. We propose a novel method for checking whether quality requirements are considered in the design stage. In this method, a technique called "spectrum analysis for quality requirements" is applied not only to requirements specifications but also to design documents. The technique enables us to derive the spectrum of a document, and quality requirements considerations in the document are numerically represented in the spectrum. We can thus objectively identify whether the considerations of quality requirements in a requirements document are adapted to its design document. To validate the method, we applied it to commercial software systems with the help of a supporting tool, and we confirmed that the method worked well.

Managing Requirements-Documents to Data

Science.gov (United States)

Orr, Kevin; Hudson, Abe

2017-01-01

Managing Requirements on long term projects like International Space Station (ISS) can go thru many phases, from initial product development to almost over 20 years of operations and sustainment. Over that time many authorized changes have been made to the requirement set, that apply to any new systems that would visit the ISS today, like commercial cargo/crew vehicles or payloads. Explore the benefits of managing requirements in a database while satisfying traditional documents needs for contracts and stakeholder/user consumption that are not tied into the database.

The need for LWR metrology standardization: the imec roughness protocol

Science.gov (United States)

Lorusso, Gian Francesco; Sutani, Takumichi; Rutigliani, Vito; van Roey, Frieda; Moussa, Alain; Charley, Anne-Laure; Mack, Chris; Naulleau, Patrick; Constantoudis, Vassilios; Ikota, Masami; Ishimoto, Toru; Koshihara, Shunsuke

2018-03-01

As semiconductor technology keeps moving forward, undeterred by the many challenges ahead, one specific deliverable is capturing the attention of many experts in the field: Line Width Roughness (LWR) specifications are expected to be less than 2nm in the near term, and to drop below 1nm in just a few years. This is a daunting challenge and engineers throughout the industry are trying to meet these targets using every means at their disposal. However, although current efforts are surely admirable, we believe they are not enough. The fact is that a specification has a meaning only if there is an agreed methodology to verify if the criterion is met or not. Such a standardization is critical in any field of science and technology and the question that we need to ask ourselves today is whether we have a standardized LWR metrology or not. In other words, if a single reference sample were provided, would everyone measuring it get reasonably comparable results? We came to realize that this is not the case and that the observed spread in the results throughout the industry is quite large. In our opinion, this makes the comparison of LWR data among institutions, or to a specification, very difficult. In this paper, we report the spread of measured LWR data across the semiconductor industry. We investigate the impact of image acquisition, measurement algorithm, and frequency analysis parameters on LWR metrology. We review critically some of the International Technology Roadmap for Semiconductors (ITRS) metrology guidelines (such as measurement box length larger than 2μm and the need to correct for SEM noise). We compare the SEM roughness results to AFM measurements. Finally, we propose a standardized LWR measurement protocol - the imec Roughness Protocol (iRP) - intended to ensure that every time LWR measurements are compared (from various sources or to specifications), the comparison is sensible and sound. We deeply believe that the industry is at a point where it is

Nondestructive evaluation of LWR spent fuel shipping casks

International Nuclear Information System (INIS)

Ballard, D.W.

1978-02-01

An analysis of nondestructve testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined. A comprehensive bibliography of current NDT methods for cask evaluation in the USA, Great Britain, Japan and West Germany was compiled for this study

Assessment of LWR piping design loading based on plant operating experience

International Nuclear Information System (INIS)

Svensson, P.O.

1980-08-01

The objective of this study has been to: (1) identify current Light Water Reactor (LWR) piping design load parameters, (2) identify significant actual LWR piping loads from plant operating experience, (3) perform a comparison of these two sets of data and determine the significance of any differences, and (4) make an evaluation of the load representation in current LWR piping design practice, in view of plant operating experience with respect to piping behavior and response to loading

Monitored Retrievable Storage System Requirements Document. Revision 1

Energy Technology Data Exchange (ETDEWEB)

1994-03-01

This Monitored Retrievable Storage System Requirements Document (MRS-SRD) describes the functions to be performed and technical requirements for a Monitored Retrievable Storage (MRS) facility subelement and the On-Site Transfer and Storage (OSTS) subelement. The MRS facility subelement provides for temporary storage, at a Civilian Radioactive Waste Management System (CRWMS) operated site, of spent nuclear fuel (SNF) contained in an NRC-approved Multi-Purpose Canister (MPC) storage mode, or other NRC-approved storage modes. The OSTS subelement provides for transfer and storage, at Purchaser sites, of spent nuclear fuel (SNF) contained in MPCs. Both the MRS facility subelement and the OSTS subelement are in support of the CRWMS. The purpose of the MRS-SRD is to define the top-level requirements for the development of the MRS facility and the OSTS. These requirements include design, operation, and decommissioning requirements to the extent they impact on the physical development of the MRS facility and the OSTS. The document also presents an overall description of the MRS facility and the OSTS, their functions (derived by extending the functional analysis documented by the Physical System Requirements (PSR) Store Waste Document), their segments, and the requirements allocated to the segments. In addition, the top-level interface requirements of the MRS facility and the OSTS are included. As such, the MRS-SRD provides the technical baseline for the MRS Safety Analysis Report (SAR) design and the OSTS Safety Analysis Report design.

Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 1. Summary: alternatives for the back of the LWR fuel cycle types and properties of LWR fuel cycle wastes projections of waste quantities; selected glossary

International Nuclear Information System (INIS)

1976-05-01

Volume I of the five-volume report contains executive and technical summaries of the entire report, background information of the LWR fuel cycle alternatives, descriptions of waste types, and projections of waste quantities. Overview characterizations of alternative LWR fuel cycle modes are also included

Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

International Nuclear Information System (INIS)

Sample, C.R.

1977-02-01

A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL

Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

Energy Technology Data Exchange (ETDEWEB)

Sample, C R [comp.

1977-02-01

A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

A Brief Assessment of North Korea's Capacities for Building an Experimental LWR

International Nuclear Information System (INIS)

Lee, Jung Hyu; An, Jin Soo

2011-01-01

On November 2010, North Korea revealed the construction site of 100 MWt (thermal) experimental LWR in the early stage with a target operation date of 2012. And they claimed that their first LWR construction project is proceeding with strictly domestic talent and resources. Introduction of LWR imposes various technical challenges, even though North Korea has experiences in the construction and management of graphite-moderated and gas-cooled reactor. So, there are doubts about whether they can successfully complete the project in time without any external support. In this paper, to estimate the fate of the LWR construction, we focused on the North Korea's capability to deal with the technical challenges which differ from those of gas-graphite reactor

LWR Spent Fuel Management for the Smooth Deployment of FBR

International Nuclear Information System (INIS)

Fukasawa, T.; Yamashita, J.; Hoshino, K.; Sasahira, A.; Inoue, T.; Minato, K.; Sato, S.

2015-01-01

Fast breeder reactors (FBR) and FBR fuel cycle are indispensable to prevent the global warming and to secure the long-term energy supply. Commercial FBR expects to be deployed from around 2050 until around 2110 in Japan by the replacement of light water reactors (LWR) after their 60 years life. The FBR deployment needs Pu (MOX) from the LWR-spent fuel (SF) reprocessing. As Japan can posses little excess Pu, its balance control is necessary between LWR-SF management (reprocessing) and FBR deployment. The fuel cycle systems were investigated for the smooth FBR deployment and the effectiveness of proposed flexible system was clarified in this work. (author)

High level waste storage tanks 242-A evaporator standards/requirement identification document

International Nuclear Information System (INIS)

Biebesheimer, E.

1996-01-01

This document, the Standards/Requirements Identification Document (S/RIDS) for the subject facility, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment. It lists those source documents from which requirements were extracted, and those requirements documents considered, but from which no requirements where taken. Documents considered as source documents included State and Federal Regulations, DOE Orders, and DOE Standards

Metal Matrix Microencapsulated Fuel Technology for LWR Applications

International Nuclear Information System (INIS)

Terrani, Kurt A.; Bell, Gary L.; Kiggans, Jim; Snead, Lance Lewis

2012-01-01

An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.

Air quality impacts due to construction of LWR waste management facilities

International Nuclear Information System (INIS)

1977-06-01

Air quality impacts of construction activities and induced housing growth as a result of construction activities were evaluated for four possible facilities in the LWR fuel cycle: a fuel reprocessing facility, fuel storage facility, fuel fabrication plant, and a nuclear power plant. Since the fuel reprocessing facility would require the largest labor force, the impacts of construction of that facility were evaluated in detail

Investigation of Nuclide Importance to Functional Requirements Related to Transport and Long-Term Storage of LWR Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

Broadhead, B.L.

1995-01-01

The radionuclide characteristics of light-water-reactor (LWR) spent fuel play key roles in the design and licensing activities for radioactive waste transportation systems, interim storage facilities, and the final repository site. Several areas of analysis require detailed information concerning the time-dependent behavior of radioactive nuclides including (1) neutron/gamma-ray sources for shielding studies, (2) fissile/absorber concentrations for criticality safety determinations, (3) residual decay heat predictions for thermal considerations, and (4) curie and/or radiological toxicity levels for materials assumed to be released into the ground/environment after long periods of time. The crucial nature of the radionuclide predictions over both short and long periods of time has resulted in an increased emphasis on thorough validation for radionuclide generation/depletion codes. Current radionuclide generation/depletion codes have the capability to follow the evolution of some 1600 isotopes during both irradiation and decay time periods. Of these, typically only 10 to 20 nuclides dominate contributions to each analysis area. Thus a quantitative ranking of nuclides over various time periods is desired for each of the analysis areas of shielding, criticality, heat transfer, and environmental dose (radiological toxicity). These rankings should allow for validation and data improvement efforts to be focused only on the most important nuclides. This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to

Expert system for estimating LWR plutonium production

International Nuclear Information System (INIS)

Sandquist, G.M.

1988-01-01

An Artificial Intelligence-Expert System called APES (Analysis of Proliferation by Expert System) has been developed and tested to permit a non proliferation expert to evaluate the capability and capacity of a specified LWR reactor and PUREX reprocessing system for producing and separating plutonium even when system information may be limited and uncertain. APES employs an expert system coded in LISP and based upon an HP-RL (Hewlett Packard-Representational Language) Expert System Shell. The user I/O interface communicates with a blackboard and the knowledge base which contains the quantitative models required to describe the reactor, selected fission product production and radioactive decay processes, Purex reprocessing and ancillary knowledge

Minutes of the 13th light water reactor pressure vessel surveillance dosimetry improvement program (LWR-PV-SDIP) meeting

International Nuclear Information System (INIS)

1984-04-01

Information is presented concerning ASTM LWR standards and program documentation; trend curves, PSF, and other test reactor metallurgical programs; PSF dosimetry and metallurgical capsule neutron and gamma environment characterization and metallurgical studies; PVS characterization program; other neutron fields; surveillance dosimetry measurement facility (SDMF) and perturbation studies; transport theory calculations; gamma field benchmarks and photo-reaction studies; and fission and non-fission sensor inventories and quality assurance

Determination of optimal LWR containment design, excluding accidents more severe than Class 8

International Nuclear Information System (INIS)

Cave, L.; Min, T.K.

1980-04-01

Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment

Development and testing of standardized procedures and reference data for LWR surveillance

International Nuclear Information System (INIS)

McElroy, W.N.

1979-02-01

The resources and talents of many national and international organizations and laboratories, both governmental and industrial, are being used to establish analysis methods for predicting the embrittlement condition of light water reactor (LWR) primary systems. The exact interrelationships and responsibilities between those developing, understanding, combining, and applying state-of-the-art technology in dosimetry, metallurgy, and fracture mechanics for reactor systems analysis are being carefully reviewed and studied. This has resulted in a more comprehensive definition of the scope of new and updated ASTM standards required for the analysis and interpretation of LWR pressure vessel surveillance results. Fifteen new and updated ASTM standards have now been identified, together with a restructuring of the main interfaces between the individual standard practices, guides, and methods. The paper briefly discusses these standards and the initial results of multi-laboratory research work involved in their validation and calibration

Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

International Nuclear Information System (INIS)

Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

1991-01-01

A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

Supplemental design requirements document, Project W026

International Nuclear Information System (INIS)

Weidert, J.R.

1993-01-01

This document supplements and extends the Functional Design Criteria, SP-W026-FDC-001, for the Waste Receiving and Processing Facility (WRAP), Module 1. It provides additional detailed requirements, summarizes key Westinghouse Hanford Company design guidance, and establishes baseline technical agreements to be used in definitive design of the WRAP-1 facility. Revision 3 of the Supplemental Design Requirements Document has been assigned an Impact Level of 3ESQ based on the content of the entire revision. The actual changes made from Revision 2 have an Impact Level of 3S and the basis for these changes was previously reviewed and approved per WHC correspondence No. 9355770

Fiscal year 1997-1998 waste information requirements document

International Nuclear Information System (INIS)

Poppiti, J.A.

1997-01-01

The Waste Information Requirements Document describes the activities of the Tank Waste Remediation System (TWRS) Characterization Project that provide characterization information on Hanford Site waste tanks. The characterization information is required to perform operations and meet the commitments of TWRS end users. These commitments are derived from the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement; the Recommendation 93-5 Implementation Plan to the Defense Nuclear Facilities Safety Board (DNFSB); and other directives as listed in Section 4.0. This Waste Information Requirement Document applies to Fiscal Years 1997 and 1998 activities. Its contents are based on the best information available in August 1997. The format and content are based on the directions of DOE-RL (Sieracki, 1997) and Fluor Daniel Hanford Incorporated (Umek, 1997). Activities, such as the revision of the Tank Characterization Technical Sampling Basis (Brown et al. 1997), the revision of the data quality objectives (DQOs), issue closures, discussions with Ecology, and management decisions may cause subsequent updates to the Waste Information Requirements Document

ISS Crew Transportation and Services Requirements Document

Science.gov (United States)

Bayt, Robert L. (Compiler); Lueders, Kathryn L. (Compiler)

2016-01-01

The ISS Crew Transportation and Services Requirements Document (CCT-REQ-1130) contains all technical, safety, and crew health medical requirements that are mandatory for achieving a Crew Transportation System Certification that will allow for International Space Station delivery and return of NASA crew and limited cargo. Previously approved on TN23183.

LIFE vs. LWR: End of the Fuel Cycle

International Nuclear Information System (INIS)

Farmer, J.C.; Blink, J.A.; Shaw, H.F.

2008-01-01

LIFE are expected to result in a more straightforward licensing process and are also expected to improve the public perception of risk from nuclear power generation, transportation of nuclear materials, and nuclear waste disposal. Waste disposal is an ongoing issue for LWRs. The conventional (once-through) LWR fuel cycle treats unburned fuel as waste, and results in the current fleet of LWRs producing about twice as much waste in their 60 years of operation as is legally permitted to be disposed of in Yucca Mountain. Advanced LWR fuel cycles would recycle the unused fuel, such that each GWe-yr of electricity generation would produce only a small waste volume compared to the conventional fuel cycle. However, the advanced LWR fuel cycle requires chemical reprocessing plants for the fuel, multiple handling of radioactive materials, and an extensive transportation network for the fuel and waste. In contrast, the LIFE engine requires only one fueling for the plant lifetime, has no chemical reprocessing, and has a single shipment of a small amount of waste per GWe-yr of electricity generation. Public perception of the nuclear option will be improved by the reduction, for LIFE engines, of the number of shipments of radioactive material per GWe-yr and the need to build multiple repositories. In addition, LIFE fuel requires neither enrichment nor reprocessing, eliminating the two most significant pathways to proliferation from commercial nuclear fuel to weapons programs

Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

International Nuclear Information System (INIS)

1977-03-01

This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document

ENHANCED CHARACTERIZATION OF THE REPOSITORY BLOCK REQUIREMENTS DOCUMENT (ECRB-RD)

International Nuclear Information System (INIS)

G.M. Teraoka

1998-01-01

This Enhanced Characterization of the Repository Block Requirements Document (ECRB-RD) provides applicable design and construction requirements for the Enhanced Characterization of the Repository Block (ECRB) East-West Drift and its associated equipment. This document also identifies the applicable requirements from the Exploratory Studies Facilities Design Requirements (ESFDR) Document (YMPICM-00 19, Revision 2, ICN- 1) for design and construction of the ECRB East-West Drift, ground support, constructor support utilities and components. These requirements have been tailored specifically for the ECRB East-West Drift design and construction. The allocated requirements for the ECRB East-West Drift are in Sections III through VI. The requirements in sections III through VI contain requirement numbers from the ESFDR, Rev 2, ICN-1 for reference back to the ESFDR. Each requirement in the ECRB-RD also identifies a trace to the Site Design and Test Requirements Document (YMP/CM-0021, Rev. 2, ICN-1) and 10CFR60 similar to the style used in the ESFDR. These traces to 10CFR60 are consistent with the 1995 version of 10CFR60 used by the SD and TRD and the ESFDR. Those ESFDR requirements statements that were technically modified are identified as such and those that were derived as part of this allocation are also identified. An activity evaluation has been performed in accordance with QAP-2-0 and has determined that the QA program is applicable to this document. Therefore, the development of this document was performed in compliance with QAP-3-5, Revision 7, Development of Technical Documents and checked and reviewed in compliance with Section 5.3. This is consistent with the IOC from R. Stambaugh to M. Lugo on the subject of ECRB-RD, Revision 1, TDPP Applicability (LV.SEI.,RMS.03/98-0 12, Dated 3/12/98). The ECRB East-West Drift includes those excavated underground openings to support enhanced characterization testing activities for the repository block and provides potential

Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation

Energy Technology Data Exchange (ETDEWEB)

None

1987-12-01

The purpose of this report, and the information contained in the associated computerized data bases, is to establish the DOE/OCRWM reference characteristics of the radioactive waste materials that may be accepted by DOE for emplacement in the mined geologic disposal system. This report provides relevant technical data for use by DOE and its supporting contractors and is not intended to be a policy document. This document is backed up by five PC-compatible data bases, written in a user-oriented, menu-driven format, which were developed for this purpose. The data bases are the LWR Assemblies Data Base; the LWR Radiological Data Base; the LWR Quantities Data Base; the LWR NFA Hardware Data Base; and the High-Level Waste Data Base. The above data bases may be ordered using the included form. An introductory information diskette can be found inside the back cover of this report. It provides a brief introduction to each of these five PC data bases. 116 refs., 18 figs., 67 tabs.

Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation

International Nuclear Information System (INIS)

1987-12-01

The purpose of this report, and the information contained in the associated computerized data bases, is to establish the DOE/OCRWM reference characteristics of the radioactive waste materials that may be accepted by DOE for emplacement in the mined geologic disposal system. This report provides relevant technical data for use by DOE and its supporting contractors and is not intended to be a policy document. This document is backed up by five PC-compatible data bases, written in a user-oriented, menu-driven format, which were developed for this purpose. The data bases are the LWR Assemblies Data Base; the LWR Radiological Data Base; the LWR Quantities Data Base; the LWR NFA Hardware Data Base; and the High-Level Waste Data Base. The above data bases may be ordered using the included form. An introductory information diskette can be found inside the back cover of this report. It provides a brief introduction to each of these five PC data bases. 116 refs., 18 figs., 67 tabs

Analysis of alternative light water reactor (LWR) fuel cycles

International Nuclear Information System (INIS)

Heeb, C.M.; Aaberg, R.L.; Boegel, A.J.; Jenquin, U.P.; Kottwitz, D.A.; Lewallen, M.A.; Merrill, E.T.; Nolan, A.M.

1979-12-01

Nine alternative LWR fuel cycles are analyzed in terms of the isotopic content of the fuel material, the relative amounts of primary and recycled material, the uranium and thorium requirements, the fuel cycle costs and the fraction of energy which must be generated at secured sites. The fuel materials include low-enriched uranium (LEU), plutonium-uranium (MOX), highly-enriched uranium-thorium (HEU-Th), denatured uranium-thorium (DU-Th) and plutonium-thorium (Pu-Th). The analysis is based on tracing the material requirements of a generic pressurized water reactor (PWR) for a 30-year period at constant annual energy output. During this time period all the created fissile material is recycled unless its reactivity worth is less than 0.2% uranium enrichment plant tails

Project W-441 cold vacuum drying facility design requirements document

International Nuclear Information System (INIS)

O'Neill, C.T.

1997-01-01

This document has been prepared and is being released for Project W-441 to record the design basis for the design of the Cold Vacuum Drying Facility. This document sets forth the physical design criteria, Codes and Standards, and functional requirements that were used in the design of the Cold Vacuum Drying Facility. This document contains section 3, 4, 6, and 9 of the Cold Vacuum Drying Facility Design Requirements Document. The remaining sections will be issued at a later date. The purpose of the Facility is to dry, weld, and inspect the Multi-Canister Overpacks before transport to dry storage

Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

Energy Technology Data Exchange (ETDEWEB)

1977-03-01

This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document.

The minimum attention plant inherent safety through LWR simplification

International Nuclear Information System (INIS)

Turk, R.S.; Matzie, R.A.

1987-01-01

The Minimum Attention Plant (MAP) is a unique small LWR that achieves greater inherent safety, improved operability, and reduced costs through design simplification. The MAP is a self-pressurized, indirect-cycle light water reactor with full natural circulation primary coolant flow and multiple once-through steam generators located within the reactor vessel. A fundamental tenent of the MAP design is its complete reliance on existing LWR technology. This reliance on conventional technology provides an extensive experience base which gives confidence in judging the safety and performance aspects of the design

Documentation and verification required for type A packaging use

Energy Technology Data Exchange (ETDEWEB)

O`Brien, J.H.

1997-07-30

This document furnishes knowledge and methods for verifying compliance with the U.S. Department of Transportation (DOT) packaging requirements for shipping Type A quantities of radioactive material. The primary emphasis is on the requirements identified in 49 CFR 173.415(a), which states, ``Each offeror of a Specification 7A package must maintain on file for at least one year after the shipment, and shall provide to DOT on request, complete documentation of tests and an engineering evaluation of comparative data showing that the construction methods, packaging design, and materials of construction comply with that specification.`` This guidance document uses a checklist to show compliance.

Documentation and verification required for type A packaging use

International Nuclear Information System (INIS)

O'Brien, J.H.

1997-01-01

This document furnishes knowledge and methods for verifying compliance with the U.S. Department of Transportation (DOT) packaging requirements for shipping Type A quantities of radioactive material. The primary emphasis is on the requirements identified in 49 CFR 173.415(a), which states, ''Each offeror of a Specification 7A package must maintain on file for at least one year after the shipment, and shall provide to DOT on request, complete documentation of tests and an engineering evaluation of comparative data showing that the construction methods, packaging design, and materials of construction comply with that specification.'' This guidance document uses a checklist to show compliance

From document to database: modernizing requirements management

International Nuclear Information System (INIS)

Giajnorio, J.; Hamilton, S.

2007-01-01

The creation, communication, and management of design requirements are central to the successful completion of any large engineering project, both technically and commercially. Design requirements in the Canadian nuclear industry are typically numbered lists in multiple documents created using word processing software. As an alternative, GE Nuclear Products implemented a central requirements management database for a major project at Bruce Power. The database configured the off-the-shelf software product, Telelogic Doors, to GE's requirements structure. This paper describes the advantages realized by this scheme. Examples include traceability from customer requirements through to test procedures, concurrent engineering, and automated change history. (author)

Assessment of management alternatives for LWR wastes. Volume 6. Cost determination of the LWR waste management routes (treatment/conditioning/packaging/transport operations)

International Nuclear Information System (INIS)

Thiels, G.M.; Kowa, S.

1993-01-01

This report deals with the cost determination of a number of schemes for the treatment, conditioning, packaging, interim storage and transport operations of LWR wastes drawn up on the basis of Belgian, French and German practices in this particular area. In addition to the general procedure elaborated for determining, actualizing and scaling of plant and transport costs associated with the various schemes, in-depth calculations of each intermediate management stage are included in this report. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

A Brief Assessment of North Korea's Capacities for Building an Experimental LWR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Hyu; An, Jin Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2011-10-15

On November 2010, North Korea revealed the construction site of 100 MWt (thermal) experimental LWR in the early stage with a target operation date of 2012. And they claimed that their first LWR construction project is proceeding with strictly domestic talent and resources. Introduction of LWR imposes various technical challenges, even though North Korea has experiences in the construction and management of graphite-moderated and gas-cooled reactor. So, there are doubts about whether they can successfully complete the project in time without any external support. In this paper, to estimate the fate of the LWR construction, we focused on the North Korea's capability to deal with the technical challenges which differ from those of gas-graphite reactor

Status of the CONTAIN computer code for LWR containment analysis

International Nuclear Information System (INIS)

Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

1983-01-01

The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

Status of the CONTAIN computer code for LWR containment analysis

International Nuclear Information System (INIS)

Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

1982-01-01

The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

Technical Development on Burn-up Credit for Spent LWR Fuel

Energy Technology Data Exchange (ETDEWEB)

Gauld, I.C.

2001-12-26

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

Technical development on burn-up credit for spent LWR fuels

International Nuclear Information System (INIS)

Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

2000-10-01

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

Technical development on burn-up credit for spent LWR fuels

Energy Technology Data Exchange (ETDEWEB)

Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-10-01

Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

Development of information management system on LWR spent fuel

International Nuclear Information System (INIS)

Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S.

2002-01-01

LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility

Development of information management system on LWR spent fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S. [KAERI, Taejon (Korea, Republic of)

2002-10-01

LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility.

Investigation of LWR environmental effect on fatigue lifetime of austenitic stainless steel component

International Nuclear Information System (INIS)

Kim, J. S.; Youm, H. K.; Jin, T. E.

1999-01-01

The fatigue lifetime of principal components in nuclear power plant is evaluated by using the design fatigue curves in ASME B and PV code during design process. However, it is inadequate to evaluate fatigue lifetime considering the LWR environmental effect by these design fatigue curves because these are presented only under atmosphere environment. Therefore, many studies are recently performed for the design fatigue curves considering LWR environmental effect and are presented that the design fatigue curves in ASME B and PV code can be non-conservative. In present paper, the limits and differences of the design fatigue curves considering environmental effect are presented. To investigate the change of fatigue lifetime according to each design fatigue curve, the CUFs for the pressurizer spray nozzle partly composed of austenitic stainless steel are calculated according to each one. Finally, if the evaluation result can not be satisfied with fatigue design requirement, the alternatives to reduce design cumulative usage factor are discussed. (author)

Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

International Nuclear Information System (INIS)

Herranz, Luis E.; Garcia, Monica; Morandi, Sonia

2013-01-01

Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have been adopted so that

B plant standards/requirements identification document (S/RID)

Energy Technology Data Exchange (ETDEWEB)

Maddox, B.S., Westinghouse Hanford

1996-07-29

This Standards/Requirements Identification Document (S/RID) set forth the Environmental Safety and Health (ES{ampersand}H) standards/requirements for the B Plant. This S/RID is applicable to the appropriate life cycle phases of design, construction,operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

Airborne radionuclide waste-management reference document

International Nuclear Information System (INIS)

Brown, R.A.; Christian, J.D.; Thomas, T.R.

1983-07-01

This report provides the detailed data required to develop a strategy for airborne radioactive waste management by the Department of Energy (DOE). The airborne radioactive materials of primary concern are tritium (H-3), carbon-14 (C-14), krypton-85 (Kr-85), iodine-129 (I-129), and radioactive particulate matter. The introductory section of the report describes the nature and broad objectives of airborne waste management. The relationship of airborne waste management to other waste management programs is described. The scope of the strategy is defined by considering all potential sources of airborne radionuclides and technologies available for their management. Responsibilities of the regulatory agencies are discussed. Section 2 of this document deals primarily with projected inventories, potential releases, and dose commitments of the principal airborne wastes from the light water reactor (LWR) fuel cycle. In Section 3, dose commitments, technologies, costs, regulations, and waste management criteria are analyzed. Section 4 defines goals and objectives for airborne waste management

Functions and requirements document for interim store solidified high-level and transuranic waste

Energy Technology Data Exchange (ETDEWEB)

Smith-Fewell, M.A., Westinghouse Hanford

1996-05-17

The functions, requirements, interfaces, and architectures contained within the Functions and Requirements (F{ampersand}R) Document are based on the information currently contained within the TWRS Functions and Requirements database. The database also documents the set of technically defensible functions and requirements associated with the solidified waste interim storage mission.The F{ampersand}R Document provides a snapshot in time of the technical baseline for the project. The F{ampersand}R document is the product of functional analysis, requirements allocation and architectural structure definition. The technical baseline described in this document is traceable to the TWRS function 4.2.4.1, Interim Store Solidified Waste, and its related requirements, architecture, and interfaces.

Functional requirements document for measuring emissions of airborne radioactive materials

International Nuclear Information System (INIS)

Glissmeyer, J.A.; Alvarez, J.L.; Hoover, M.D.; Newton, G.C.; McFarland, A.R.; Rodgers, J.C.

1994-11-01

This document states the general functional requirements for systems and procedures for measuring emissions of airborne radioactive materials from facilities administered by the Westinghouse Hanford Company (WHC). The following issues are addressed in this document: lg-bullet definition of the program objectives lg-bullet selection of the overall approach to collecting the samples lg-bullet sampling equipment design lg-bullet sampling equipment maintenance and quality assurance issues. The following issues are not addressed in this document: lg-bullet air sampling in work areas or containments lg-bullet selection of specific on-line sample monitoring instrumentation lg-bullet analyzing collected samples lg-bullet reporting and interpreting results. The document provides equipment design guidance that is performance based rather than prescriptive. Locations from which samples are obtained should exhibit mixing of the contaminants with the airstream and acceptable air flow characteristics. Sample collection equipment and effluent and sample flow elements should meet defined performance standards. Quality control and assurance requirements specific to sample collection, equipment inspection, and calibration are presented. Key sample collection performance requirements are summarized in Section 5.4. The intent of this document is to assist WHC in demonstrating a high quality of air emission measurements with verified system performance based on documented system design, testing, inspection, and maintenance

Outline of Swedish activities on LWR fuel

Energy Technology Data Exchange (ETDEWEB)

Grounes, M [Studsvik Nuclear, Nykoeping (Sweden); Roennberg, G [OKG AB (Sweden)

1997-12-01

The presentation outlines the Swedish activities on LWR fuel and considers the following issues: electricity production; performance of operating nuclear power plants; nuclear fuel cycle and waste management; research and development in nuclear field. 4 refs, 4 tabs.

Requirements for the data transfer during the examination of design documentation

Directory of Open Access Journals (Sweden)

Karakozova Irina

2017-01-01

Full Text Available When you transfer the design documents to the examination office, number of incompatible electronic documents increases dramatically. The article discusses the way to solve the problem of transferring of the text and graphic data of design documentation for state and non-state expertise, as well as verification of estimates and requirement management. The methods for the recognition of the system elements and requirements for the transferring of text and graphic design documents are provided. The need to use the classification and coding of various elements of information systems (structures, objects, resources, requirements, contracts, etc. in data transferring systems is indicated separately. The authors have developed a sequence of document processing and transmission of data during the examination, and propose a language for describing the construction of the facility, taking into account the classification criteria of the structures and construction works.

Data needs for long-term dry storage of LWR fuel. Interim report

International Nuclear Information System (INIS)

Einziger, R.E.; Baldwin, D.L.; Pitman, S.G.

1998-04-01

The NRC approved dry storage of spent fuel in an inert environment for a period of 20 years pursuant to 10CFR72. However, at-reactor dry storage of spent LWR fuel may need to be implemented for periods of time significantly longer than the NRC's original 20-year license period, largely due to uncertainty as to the date the US DOE will begin accepting commercial spent fuel. This factor is leading utilities to plan not only for life-of-plant spent-fuel storage during reactor operation but also for the contingency of a lengthy post-shutdown storage. To meet NRC standards, dry storage must (1) maintain subcriticality, (2) prevent release of radioactive material above acceptable limits, (3) ensure that radiation rates and doses do not exceed acceptable limits, and (4) maintain retrievability of the stored radioactive material. In light of these requirements, this study evaluates the potential for storing spent LWR fuel for up to 100 years. It also identifies major uncertainties as well as the data required to eliminate them. Results show that the lower radiation fields and temperatures after 20 years of dry storage promote acceptable fuel behavior and the extension of storage for up to 100 years. Potential changes in the properties of dry storage system components, other than spent-fuel assemblies, must still be evaluated

Fabrication data package for HEDL dosimetry in the ORNL Poolside Facility: LWR Pressure Vessel Mock-up irradiation

International Nuclear Information System (INIS)

Lippincott, E.P.; McElroy, W.N.; Kellogg, L.S.; Gold, R.; Guthrie, G.L.; Ruddy, F.H.; Ulseth, J.A.

1981-09-01

This document provides a complete description of the HEDL dosimetry inserted in the metallurgical specimen irradiation in the LWR Pressure Vessel Mock-up at the Oak Ridge Reactor Poolside Facility (PSF). This experiment is being conducted under the Nuclear Regulatory Commission sponsored program on Surveillance Dosimetry Improvement. The irradiation started April 1980 with recovery of the 2 x 10 19 (nominal fluence with E > 1 MeV) capsule in September 1980, the 4 x 10 19 surveillance capsule in November 1981 and the pressure vessel and void box capaules about August 1982

AFCI : Co-extraction impacts on LWR and fast reactor fuel cycles

International Nuclear Information System (INIS)

Taiwo, T. A.; Szakalay, F. J.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

2007-01-01

A systematic investigation of the impact of the co-extraction COEXTM process on reactor performance has been performed. The proliferation implication of the process was also evaluated using the critical mass, radioactivity, decay heat and neutron and gamma source rates and gamma doses as indicators. The use of LWR-spent-uranium-based MOX fuel results in a higher initial plutonium content requirement in an LWR MOX core than if natural uranium based MOX fuel is used (by about 1%); the plutonium for both cases is derived from the spent LWR spent fuel. More transuranics are consequently discharged in the spent fuel of the MOX core. The presence of U-236 in the initial fuel was also found to result in higher content of Np-237 in the spent MOX fuel and less consumption of Pu-238 and Am-241 in the MOX core. The higher quantities of Np-237 (factor of 5), Pu-238 (20%) and Am-241 (14%) decrease the effective repository utilization, relative to the use of natural uranium in the PWR MOX core. Additionally, the minor actinides continue to accumulate in the fuel cycle, even if the U-Pu co-extraction products are continuously recycled in the PWR cores, and thus a solution is required for the minor actinides. The utilization of plutonium derived from LWR spent fuel versus weapons-grade plutonium for the startup core of a 1,000 MWT advanced burner fast reactor (ABR) increases the TRU content by about 4%. Differences are negligible for the equilibrium recycle core. The impact of using reactor spent uranium instead of depleted uranium was found to be relatively smaller in the fast reactor (TRU content difference less than 0.4%). The critical masses of the co-extraction products were found to be higher than that of weapons-grade plutonium and the decay heat and radiation sources of the materials (products) were also found to be generally higher than that of weapons-grade plutonium (WG-Pu) in the transuranics content range of 0.1 to 1.0 in the heavy-metal. The magnitude of the

Probabilistic safety assessment of LWR containment systems performance. Report of principal working group n.5 on risk assessment

International Nuclear Information System (INIS)

Holloway, N.J.; Harper, F.T.; Bellard, S.W.

1992-01-01

This report reviews current approaches to PSA of LWR containment systems performance. It is based on a variety of recent PSA reports which deal with Level-2 PSA. The report is a summary of recent state-of-the-art containment analysis and is intended to assist analysts in their selection of the most appropriate methods of extending Level-1 plant safety evaluations into Level-2 assessments of the containment performance. The document is primarily concerned with the performance of the containment as an engineered system rather than with the source terms consequent upon its failure. It is addressed mainly to the performance of large dry PWR containments, with a secondary emphasis on other containment types. After explaining the purposes of these analyses, a survey of LWR containment analysis options is presented: direct approaches using containment event tree construction, indirect approaches based on previous PSAs, alternative and novel approaches. The selection process is then described, followed by conclusions on their suitability for various cases: accident management, research prioritization, identifying design weaknesses, specific issue resolution, modelling physical reality, etc.

Principles of MONJU maintenance. Characteristic of MONJU maintenance and reflection of LWR maintenance experience to FBR

International Nuclear Information System (INIS)

Nakai, Satoru; Nishio, Ryuichi; Uchihashi, Masaya; Kaneko, Yoshihisa; Yamashita, Hironobu; Yamaguchi, Atsunori; Aoki, Takayuki

2014-01-01

A sodium cooled fast breeder reactor (FBR) has unique systems and components and different degradation mechanism from light water reactor (LWR) so that need to establish maintenance technology in accordance with its features. The examination of the FBR maintenance technology is carried out in the special committee for considering the maintenance for Monju established in the Japan Society of Maintenology (JSM). As a result of the study such as extraction of Monju maintenance feature, maintenance technology benchmark between Monju and LWR components and survey of LWR maintenance experience, it is clear that principles of maintenance are same as LWR, necessity of LWR maintenance experience reflection and points to be considered in Monju maintenance. The road map to establish a FBR maintenance technology in the technical aspect became clear and it is vital to acquire operation and maintenance experience of the plant to implement this road map, and to establish a fast reactor maintenance. (author)

FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

Directory of Open Access Journals (Sweden)

WEON-JU KIM

2013-08-01

Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

Preliminary design requirements document (DRD) for Project W-236B, ''Initial Pretreatment Module''

International Nuclear Information System (INIS)

Swanson, L.M.

1995-01-01

The scope of this Design Requirements Document (DRD) is to identify and define the functions, with associated requirements, which must be performed to separate Hanford Site tank waste supernatants into low-level and high-level fractions. This documents sets forth function requirements, performance requirements, and design constraints necessary to begin conceptual design for the Initial Pretreatment Module (IPM). System and physical interfaces between the IPM project and the Tank Waste Remediation System (TWRS) are identified. The constraints, performance requirements, and transfer of information and data across a technical interface will be documented in an Interface Control Document. Supplemental DRDs will be prepared to provide more detailed requirements specific to systems described in the DRD

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

Directory of Open Access Journals (Sweden)

Saas Laurent

2017-01-01

Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…. During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.

Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

Energy Technology Data Exchange (ETDEWEB)

Herranz, Luis E., E-mail: luisen.herranz@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Garcia, Monica, E-mail: monica.gmartin@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Morandi, Sonia, E-mail: sonia.morandi@rse-web.it [Nuclear and Industrial Plant Safety Team, Power Generation System Department, RSE, via Rubattino 54, 20134 Milano (Italy)

2013-12-15

Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have

Hanford Tanks Initiative requirements and document management process guide

International Nuclear Information System (INIS)

Schaus, P.S.

1998-01-01

This revision of the guide provides updated references to project management level Program Management and Assessment Configuration Management activities, and provides working level directions for submitting requirements and project documentation related to the Hanford Tanks Initiative (HTI) project. This includes documents and information created by HTI, as well as non-HTI generated materials submitted to the project

Generalized perturbation theory for LWR depletion analysis and core design applications

International Nuclear Information System (INIS)

White, J.R.; Frank, B.R.

1986-01-01

A comprehensive time-dependent perturbation theory formulation that includes macroscopic depletion, thermal-hydraulic and poison feedback effects, and a criticality reset mechanism is developed. The methodology is compatible with most current LWR design codes. This new development allows GTP/DTP methods to be used quantitatively in a variety of realistic LWR physics applications that were not possible prior to this work. A GTP-based optimization technique for incore fuel management analyses is addressed as a promising application of the new formulation

Dual-purpose LWR supplying heat for desalination

International Nuclear Information System (INIS)

Waplington, G.; Fitcher, H.

1977-01-01

A number of desalination processes are at present in various stages of development but distillation is the only serious choice for a large-scale project. The distillation process temperature requirement is low compared with the temperature of steam normally delivered to the turbine in a power generation plant. This gives the possibility for combining the functions of electricity generation with water distillation. The brine heater of the multi-stage flash distillation plant can be supplied with steam after partial expansion through a turbine. Such an arrangement allows the use of a standard nuclear steam supply system and makes fuller use of the energy output than would either single purpose role. The LWR represents a safe, reliable and economic system, and is easily able to provide heat of a quality adequate for the desalination process. (M.S.)

Advanced light water reactor utility requirements document: Volume 1--ALWR policy and summary of top-tier requirements

International Nuclear Information System (INIS)

Anon.

1990-01-01

The U.S. utilities are leading an industry wide effort to establish the technical foundation for the design of the Advanced Light Water Reactor (ALWR). This effort, the ALWR Program, is being managed for the U.S. electric utility industry by the Electric Power Research Institute (EPRI) and includes participation and sponsorship of several international utility companies and close cooperation with the U.S. Department of Energy (DOE). The cornerstone of the ALWR Program is a set of utility design requirements which are contained in the ALWR Requirements Document. The purpose of the Requirement Document is to present a clear, complete statement of utility desires for their next generation of nuclear plants. The Requirements Document covers the entire plant up to the grid interface. It therefore is the basis for an integrated plant design, i.e., nuclear steam supply system and balance of plant, and it emphasizes those areas which are most important to the objective of achieving an ALWR which is excellent with respect to safety, performance, constructibility, and economics. The document applies to both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The Requirements Document is organized in three volumes. Volume 1 summarizes AlWR Program policy statements and top-tier requirements. The top-tier design requirements are categorized by major functions, including safety and investment protection, performance, and design process and constructibility. There is also a set of general design requirements, such as simplification and proven technology, which apply broadly to the ALWR design, and a set of economic goals for the ALWR program. The top-tier design requirements are described further in Volume 1 and are formally invoked as requirements in Volumes 2 and 3

Design requirements document for Project W-465, immobilized low-activity waste interim storage

International Nuclear Information System (INIS)

Burbank, D.A.

1998-01-01

The scope of this Design Requirements Document (DRD) is to identify the functions and associated requirements that must be performed to accept, transport, handle, and store immobilized low-activity waste (ILAW) produced by the privatized Tank Waste Remediation System (TWRS) treatment contractors. The functional and performance requirements in this document provide the basis for the conceptual design of the TWRS ILAW Interim Storage facility project and provides traceability from the program level requirements to the project design activity. Technical and programmatic risk associated with the TWRS planning basis are discussed in the Tank Waste Remediation System Decisions and Risk Assessment (Johnson 1994). The design requirements provided in this document will be augmented by additional detailed design data documented by the project

Assesment On The Possibility To Modify Fabrication Equipment For Fabrication Of HWR And LWR Fuel Elements

International Nuclear Information System (INIS)

Tri-Yulianto

1996-01-01

Based on TOR BATAN for PELITA VI. On of BATAN program in the fuel element production technology section is the acquisition of the fuel element fabrication technology for research reactor as well as power reactor. The acquisition can be achieved using different strategies, e.g. by utilizing the facility owned for research and development of the technology desired or by transferring the technology directly from the source. With regards to the above, PEBN through its facility in BEBE has started the acquisition of the fuel element fabrication technology for power reactor by developing the existing equipment initially designed to fabricate HWR Cinere fuel element. The development, by way of modifying the equipment, is intended for the production of HWR (Candu) and LWR (PWR and BWR) fuel elements. To achieve above objective, at the early stage of activity, an assesment on the fabrication equipment for pelletizing, component production and assembly. The assesment was made by comparing the shape and the size of the existing fuel element with those used in the operating reactors such as Candu reactors, PWR and BWR. Equipment having the potential to be modified for the production of HWR fuel elements are as followed: For the pelletizing equipment, the punch and dies can be used of the pressing machine for making green pellet can be modified so that different sizes of punch and dies can be used, depending upon the size of the HWR and LWR pellets. The equipment for component production has good potential for modification to produce the HWR Candu fuel element, which has similar shape and size with those of the existing fuel element, while the possibility of producing the LWR fuel element component is small because only a limited number of the required component can be made with the existing equipment. The assembly equipment has similar situation whit that of the component production, that is, to assemble the HWR fuel element modification of few assembly units very probable

Future CANDU nuclear power plant design requirements document executive summary

International Nuclear Information System (INIS)

Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; S. A. Usmani

1996-03-01

The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new

Development of top nozzle for Korean standard LWR fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, S. K.; Kim, I. K.; Choi, K. S.; Kim, Y. H.; Lee, J. N.; Kim, H. K. [KNFC, Taejon (Korea, Republic of)

2001-10-01

Performance evaluation was executed for each component and its assembly for the deduced Top Nozzles to develop the new Top Nozzle for LWR. This new Top Nozzle is composed of the optimum components among the derived Top Nozzles that have been evaluated in the viewpoint of structural integrity, simpleness of dismantle and assembly, manufacturability etc. In this study, the developed Top Nozzle satisfied all the related design criteria. In special, it makes fuel repair time reduced by assembling and disassembling itself as one body, and improves Fuel Assembly holddown ability by revising the design parameters of its spring and the structural integrity through the betterment of its geometrical shpae of Flange and Holddown Plate as compared with the existing LWR Top Nozzles.

Feasibility assessment of the once-through thorium fuel cycle for the PTVM LWR concept

International Nuclear Information System (INIS)

Rachamin, R.; Fridman, E.; Galperin, A.

2015-01-01

Highlights: • The PTVM LWR is an innovation reactor concept operating in a “breed & burn” mode. • An advanced once-through thorium fuel cycle for the PTVM LWR concept is proposed. • The PTVM LWR concept makes use of a seed-blanket geometry. • A novel fuel management scheme based on two separate fuel flow routes is analyzed. • The analysis indicates a potential for utilizing the fuel in an efficient manner. - Abstract: This paper investigates the feasibility of a once-through thorium fuel cycle for the novel reactor-design concept named the pressure tube light water reactor with variable moderator control (PTVM LWR). The PTVM LWR operates in a “breed & burn” mode, which makes it an attractive system for utilizing thorium fuel in a once-through mode. The “breed & burn” mode can emphasize the in situ generation as well as incineration of 233 U, which are the basic foundations of the once-through thorium fuel cycle. The PTVM LWR concept makes use of a seed–blanket geometry, whereby the core is divided into separated regions of thorium-based fuel channel assemblies (blanket) and low-enriched uranium (LEU) based fuel channel assemblies (seed). A novel fuel in-core management scheme based on two separate fuel flow routes (i.e., seed route and blanket route) is proposed and analyzed. Neutronic performance analysis indicates that the proposed novel fuel in-core management scheme has the potential to utilize both LEU- and thorium-based fuel in an efficient manner. The once-through thorium cycle, presented and discussed in this paper, provide interesting research leads and can serve as a bridge between current LEU-based fuel cycles and a thorium fuel cycle based on recycling of 233 U

LWR and HTGR coolant dynamics: the containment of severe accidents

International Nuclear Information System (INIS)

Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

1983-07-01

This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

Development of training simulator for LWR

International Nuclear Information System (INIS)

Sureshbabu, R.M.

2009-01-01

A full-scope training simulator was developed for a light water reactor (LWR). This paper describes how the development evolved from a desktop simulator to the full-scope training simulator. It also describes the architecture and features of the simulator including the large number of failures that it simulates. The paper also explains the three-level validation tests that were used to qualify the training simulator. (author)

Preliminary Design Requirements Document for Project W-314

Energy Technology Data Exchange (ETDEWEB)

MCGREW, D.L.

2000-04-27

This document sets forth functional requirements, performance requirements, and design constraints for the tank farm systems elements identified in Section 3.1 of this document. These requirements shall be used to develop the Design Requirements Baseline for those system elements. System Overview--The tank farm system at Hanford Site currently consists of 149 single shell tanks and 28 double shell tanks with associated facilities and equipment, located in 18 separate groupings. Each grouping is known as a tank farm. They are located in the areas designated as 200 West and 200 East. Table 1-1 shows the number of tanks in each farm. The farms are connected together through a transfer system consisting of piping, diversion boxes, Double Contained Receiver Tanks (DCRT) and other miscellaneous facilities and elements. The tank farm system also connects to a series of processing plants which generate radioactive and hazardous wastes. The primary functions of the tank farm system are to store, transfer, concentrate, and characterize radioactive and hazardous waste generated at Hanford, until the waste can be safely retrieved, processed and dispositioned. The systems provided by Project W-314 support the store and transfer waste functions. The system elements to be upgraded by Project W-314 are identified in Section 3.1.

Preliminary Design Requirements Document for Project W-314

International Nuclear Information System (INIS)

MCGREW, D.L.

2000-01-01

This document sets forth functional requirements, performance requirements, and design constraints for the tank farm systems elements identified in Section 3.1 of this document. These requirements shall be used to develop the Design Requirements Baseline for those system elements. System Overview--The tank farm system at Hanford Site currently consists of 149 single shell tanks and 28 double shell tanks with associated facilities and equipment, located in 18 separate groupings. Each grouping is known as a tank farm. They are located in the areas designated as 200 West and 200 East. Table 1-1 shows the number of tanks in each farm. The farms are connected together through a transfer system consisting of piping, diversion boxes, Double Contained Receiver Tanks (DCRT) and other miscellaneous facilities and elements. The tank farm system also connects to a series of processing plants which generate radioactive and hazardous wastes. The primary functions of the tank farm system are to store, transfer, concentrate, and characterize radioactive and hazardous waste generated at Hanford, until the waste can be safely retrieved, processed and dispositioned. The systems provided by Project W-314 support the store and transfer waste functions. The system elements to be upgraded by Project W-314 are identified in Section 3.1

Advanced LWR Nuclear Fuel Cladding Development

International Nuclear Information System (INIS)

Bragg-Sitton, S.; Griffith, G.

2012-01-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

Detector Control System for an LHC experiment - User Requirements Document

CERN Document Server

CERN. Geneva

1997-01-01

The purpose of this document is to provide the user requirements for a detector control system kernel for the LHC experiments following the ESA standard PSS-05 [1]. The first issue will be used to provide the basis for an evaluation of possible development philosophies for a kernel DCS. As such it will cover all the major functionality but only to a level of detail sufficient for such an evaluation to be performed. Many of the requirements are therefore intentionally high level and generic, and are meant to outline the functionality that would be required of the kernel DCS, but not yet to the level of the detail required for implementation. The document is also written in a generic fashion in order not to rule out any implementation technology.

Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-08-29

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U₃Si₂) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U₃Si₂ at LWR conditions. The fission gas behavior of U₃Si₂ can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U₃Si₂ for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U₃Si₂ at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U₃Si₂ as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

International Nuclear Information System (INIS)

Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

1998-01-01

This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

Energy profit ratio on LWR by uranium recycles

International Nuclear Information System (INIS)

Amano, Osamu; Uno, Takeki; Matsushima, Jun

2009-01-01

Energy profit ratio is defined as the ratio of output energy/input system total energy. In case of electric power generation, input energy is a total for fuel such as uranium mining and enrichment, fuel transportation, build nuclear power plant, M and O and for disposal waste and decommission of reactor vessel. Output energy is the total electricity on LWR during the plant life. EPR on both PWR and BWR is high value using gas centrifuge enrichment compared other type of electric power generation such as a thermal power, a hydraulic power, a wind power and a photovoltaic power. How is the EPR on LWR by MOX? We need understanding the energy of reprocessing spent fuel, MOX fuel fabrication, low level waste disposal and high level radioactive glass disposal. As we show the material balance for two cases, the first is the case of long term storage and reprocessing before FBR, the second is the MOX fuel cycle on LWR plant. The MOX fuel recycle is better EPR value rather than the case of long term storage and reprocessing before FBR (LTSRBF). At the gaseous diffusion enrichment case, MOX fuel recycle has 15 to 18% higher EPR value than LTSRBF. At the gas centrifuge enrichment case the MOX fuel recycle has 17 to 18 higher EPR value than LTSRBF. MOX fuel recycle decreases the uranium mining and refine mass, enrichment separative work and the spent fuel interim storage. It tells us the MOX fuel recycle is good way from view of EPR. (author)

Lawrence Livermore National Laboratory Emergency Response Capability Baseline Needs Assessment Requirement Document

Energy Technology Data Exchange (ETDEWEB)

Sharry, J A

2009-12-30

This revision of the LLNL Fire Protection Baseline Needs Assessment (BNA) was prepared by John A. Sharry, LLNL Fire Marshal and LLNL Division Leader for Fire Protection and reviewed by Martin Gresho, Sandia/CA Fire Marshal. The document follows and expands upon the format and contents of the DOE Model Fire Protection Baseline Capabilities Assessment document contained on the DOE Fire Protection Web Site, but only address emergency response. The original LLNL BNA was created on April 23, 1997 as a means of collecting all requirements concerning emergency response capabilities at LLNL (including response to emergencies at Sandia/CA) into one BNA document. The original BNA documented the basis for emergency response, emergency personnel staffing, and emergency response equipment over the years. The BNA has been updated and reissued five times since in 1998, 1999, 2000, 2002, and 2004. A significant format change was performed in the 2004 update of the BNA in that it was 'zero based.' Starting with the requirement documents, the 2004 BNA evaluated the requirements, and determined minimum needs without regard to previous evaluations. This 2010 update maintains the same basic format and requirements as the 2004 BNA. In this 2010 BNA, as in the previous BNA, the document has been intentionally divided into two separate documents - the needs assessment (1) and the compliance assessment (2). The needs assessment will be referred to as the BNA and the compliance assessment will be referred to as the BNA Compliance Assessment. The primary driver for separation is that the needs assessment identifies the detailed applicable regulations (primarily NFPA Standards) for emergency response capabilities based on the hazards present at LLNL and Sandia/CA and the geographical location of the facilities. The needs assessment also identifies areas where the modification of the requirements in the applicable NFPA standards is appropriate, due to the improved fire protection

Effects of Listening While Reading (LWR on Swahili Reading Fluency and Comprehension

Directory of Open Access Journals (Sweden)

Filipo Lubua

2016-10-01

Full Text Available A number of studies have examined the contribution of technology in teaching such languages as English, French, and Spanish, among many others. Contrarily, most LCTL’s, have received very little attention. This study investigates if listening while reading (LWR may expedite Swahili reading fluency and comprehension. The study employed the iBook Author tool to create weekly mediated and interactive reading texts, with comprehension exercises, which were eventually used to collect descriptive and qualitative data from four Elementary Swahili students. Participants participated in a seven week reading program, which provided them with some kind of directed self-learning, and met with the instructor for at least 30 minutes every week for observation and more reading activities. The teacher recorded their reading scores, and a number of themes on how LWR influenced reading fluency and comprehension are discussed here. It shows that participants have a positive attitude towards LWR and they suggest it for all the reading classes.

Civilian Radioactive Waste Management System Requirements Document (CRP)

International Nuclear Information System (INIS)

C.A. Kouts

2006-01-01

The CRD addresses the requirements of Department of Energy (DOE) Order 413.3-Change 1, ''Program and Project Management for the Acquisition of Capital Assets'', by providing the Secretarial Acquisition Executive (Level 0) scope baseline and the Program-level (Level 1) technical baseline. The Secretarial Acquisition Executive approves the Office of Civilian Radioactive Waste Management's (OCRWM) critical decisions and changes against the Level 0 baseline; and in turn, the OCRWM Director approves all changes against the Level 1 baseline. This baseline establishes the top-level technical scope of the CRMWS and its three system elements, as described in section 1.3.2. The organizations responsible for design, development, and operation of system elements described in this document must therefore prepare subordinate project-level documents that are consistent with the CRD. Changes to requirements will be managed in accordance with established change and configuration control procedures. The CRD establishes requirements for the design, development, and operation of the CRWMS. It specifically addresses the top-level governing laws and regulations (e.g., ''Nuclear Waste Policy Act'' (NWPA), 10 Code of Federal Regulations (CFR) Part 63, 10 CFR Part 71, etc.) along with specific policy, performance requirements, interface requirements, and system architecture. The CRD shall be used as a vehicle to incorporate specific changes in technical scope or performance requirements that may have significant program implications. Such may include changes to the program mission, changes to operational capability, and high visibility stakeholder issues. The CRD uses a systems approach to: (1) identify key functions that the CRWMS must perform, (2) allocate top-level requirements derived from statutory, regulatory, and programmatic sources, and (3) define the basic elements of the system architecture and operational concept. Project-level documents address CRD requirements by further

Evaluation of management alternatives for LWR hulls and caps

International Nuclear Information System (INIS)

Chaudon, L.; Mehling, O.; Cecille, L.; Thiels, G.; Kowa, S.

1993-01-01

Hulls and caps resulting from the reprocessing of LWR spent fuels represent one of the major sources of alpha-bearing solid waste generated during the nuclear fuel cycle. The Commission of the European Communities has undertaken considerable R and D efforts on the development of advanced treatment and conditioning methods for this type of waste. In view of the encouraging results achieved, the Commission launched a theoretical assessment study on cladding waste management. Six practical or potential schemes were identified and elaborated: direct cementation, decontamination prior to cementation, rolling before cementation, rolling followed by embedding in graphite, compaction, and melting in a cold crucible. The economic aspects of each management option were also investigated. This included the assessment of the plant (treatment, conditioning and interim storage), transport and disposal costs. Further consideration will be required to define the best management option for 'cap' wastes. Transport and disposal costs will also require further analysis from an industrial standpoint

Vectorization of LWR transient analysis code RELAP5/MOD1 and its effect

International Nuclear Information System (INIS)

Ishiguro, Misako; Harada, Hiroo; Shinozawa, Naohisa; Naraoka, Ken-itsu

1985-03-01

The RELAP5/MOD1 is a large thermal-hydraulic code to analyze LWR LOCA and non-LOCA transients. The code originally was designed for use on a CDC Cyber-176. This report documents vectorization of the RELAP5/MOD1 code conducted for the purpose of efficient use of VP-100 (peak speed 250 MFLOPS, clock period 7.5 ns) at the JAERI. The code was vectorized using the junction and volume level parallelisms in the hydrodynamic calculations, and the heat-structure and heat-mesh level in the heat conduction calculations. The vectorized version runs as much as 2.4 to 2.8 times faster than the original scalar version, while the speedup ratio is dependent on the number of spactial cells included in the problem. (author)

Quality assurance in the course of fabrication of LWR fuel

International Nuclear Information System (INIS)

Dressler, G.; Perry, J.A.

1982-01-01

A high quality level of LWR fuel elements can only be assured by a system of Quality Assurance measures purposefully designed, balanced, and appropriately applied. This includes application of and the appropriate balance between both system and product oriented measures. A prerequisite to the establishment of these measures is a precise analysis of the various influences of the individual process steps on the quality characteristics of the starting materials, semi-finished and finished products. In addition, these characteristics require classification criteria relative to their significance. The described classification is used to establish sampling plans and to disposition non-conformances. The EXXON Nuclear Quality Assurance system which is based on these principles is described and illustrated with some examples. (orig.)

Preliminary Study for Radioactivity Evaluation of MSR compared with LWR

International Nuclear Information System (INIS)

Lee, Geun Hyeong; Kim, Hee Reyoung

2014-01-01

LWR uses fuel as 235 U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile 233 U when 232 Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster

Preliminary Study for Radioactivity Evaluation of MSR compared with LWR

Energy Technology Data Exchange (ETDEWEB)

Lee, Geun Hyeong; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-05-15

LWR uses fuel as {sup 235}U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile {sup 233}U when {sup 232}Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster.

LWR-PV Surveillance Dosimetry Improvement Program review graphics

International Nuclear Information System (INIS)

McElroy, W.N.; Gold, R.; Gutherie, G.L.

1979-10-01

A primary objective of the multilaboratory program is to prepare an updated and improved set of dosimetry, damage correlation, and the associated reactor analysis ASTM standards for LWR-PV irradiation surveillance programs. Supporting this objective are a series of analytical and experimental validation and calibration studies in Benchmark Neutron Fields, reactor Test Regions, and operating power reactor Surveillance Positions. These studies will establish and certify the precision and accuracy of the measurement and predictive methods which are recommended for use in these standards. Consistent and accurate measurement and data analysis techniques and methods, therefore, will have been developed and validated along with guidelines for required neutron field calculations that are used to (1) correlate changes in material properties with the characteristics of the neutron radiation field and (2) predict pressure vessel steel toughness and embrittlement from power reactor surveillance data

Advanced hybrid process with solvent extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR

International Nuclear Information System (INIS)

Fujita, Reiko; Mizuguchi, Koji; Fuse, Kouki; Saso, Michitaka; Utsunomiya, Kazuhiro; Arie, Kazuo

2008-01-01

Toshiba has been proposing a new fuel cycle concept of a transition from LWR to FBR. The new fuel cycle concept has better economical process of the LWR spent fuel reprocessing than the present Purex Process and the proliferation resistance for FBR cycle of plutonium with minor actinides after 2040. Toshiba has been developing a new Advanced Hybrid Process with Solvent Extraction and Pyrochemical process of spent fuel reprocessing for LWR to FBR. The Advanced Hybrid Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high pure uranium for LWR fuel and the pyro-chemical process in molten salts of impure plutonium recovery with minor actinides for metallic FBR fuel, which is the FBR spent fuel recycle system after FBR age based on the electrorefining process in molten salts since 1988. The new Advanced Hybrid Process enables the decrease of the high-level waste and the secondary waste from the spent fuel reprocessing plants. The R and D costs in the new Advanced Hybrid Process might be reduced because of the mutual Pyro-chemical process in molten salts. This paper describes the new fuel cycle concept of a transition from LWR to FBR and the feasibility of the new Advanced Hybrid Process by fundamental experiments. (author)

In-core materials testing under LWR conditions in the Halden reactor

International Nuclear Information System (INIS)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

2002-01-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

In-core materials testing under LWR conditions in the Halden reactor

Energy Technology Data Exchange (ETDEWEB)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

2002-07-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

7 CFR 1488.8 - Documents required after delivery.

Science.gov (United States)

2010-01-01

... 7 Agriculture 10 2010-01-01 2010-01-01 false Documents required after delivery. 1488.8 Section... delivery. (a) CCC will purchase an exporter's account receivable only if the Treasurer, Commodity Credit... or Assistant Treasurer, CCC, after date of delivery of commodities exported or to be exported under...

Development of LWR fuel performance code FEMAXI-6

International Nuclear Information System (INIS)

Suzuki, Motoe

2006-01-01

LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

Cost optimization of long-cycle LWR operation

International Nuclear Information System (INIS)

Handwerk, C.S.; Driscoll, M.J.; McMahon, M.V.; Todreas, N.E.

1997-01-01

The continuing emphasis on improvement of plant capacity factor, as a major means to make nuclear energy more cost competitive in the current deregulatory environment, motivates heightened interest in long intra-refueling intervals and high burnup in LWR units. This study examines the economic implications of these trends, to determine the envelope of profitable fuel management tactics. One batch management is found to be significantly more expensive than two-batch management. Parametric studies were carried out varying the most important input parameters. If ultra-high burnup can be achieved, then n = 3 or even n = 4 management may be preferable. For n = 1 or 2, economic performance declines at higher burnups, hence providing no great incentive for moving further in that direction. Values for n > 2 are also attractive because, for a given burnup target, required enrichment decreases as n increases. This study was limited to average batch burnups below 60,000 MWd/MT

Westinghouse Hanford Company (WHC) standards/requirements identification document (S/RID)

Energy Technology Data Exchange (ETDEWEB)

Bennett, G.L.

1996-03-15

This Standards/Requirements Identification Document (S/RID) set forth the Environmental Safety and Health (ES&H) standards/requirements for Westinghouse Hanford Company Level Programs, where implementation and compliance is the responsibility of these organizations. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

IDC Re-Engineering Phase 2 System Requirements Document V1.3.

Energy Technology Data Exchange (ETDEWEB)

Harris, James M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Burns, John F. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Satpathi, Meara Allena [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-12-01

This System Requirements Document (SRD) defines waveform data processing requirements for the International Data Centre (IDC) of the Comprehensive Nuclear Test Ban Treaty Organization (CTBTO). The IDC applies, on a routine basis, automatic processing methods and interactive analysis to raw International Monitoring System (IMS) data in order to produce, archive, and distribute standard IDC products on behalf of all States Parties. The routine processing includes characterization of events with the objective of screening out events considered to be consistent with natural phenomena or non-nuclear, man-made phenomena. This document does not address requirements concerning acquisition, processing and analysis of radionuclide data but includes requirements for the dissemination of radionuclide data and products.

IDC Re-Engineering Phase 2 System Requirements Document Version 1.4

Energy Technology Data Exchange (ETDEWEB)

Harris, James M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Burns, John F. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Satpathi, Meara Allena [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-01-01

This System Requirements Document (SRD) defines waveform data processing requirements for the International Data Centre (IDC) of the Comprehensive Nuclear Test Ban Treaty Organization (CTBTO). The IDC applies, on a routine basis, automatic processing methods and interactive analysis to raw International Monitoring System (IMS) data in order to produce, archive, and distribute standard IDC products on behalf of all States Parties. The routine processing includes characterization of events with the objective of screening out events considered to be consistent with natural phenomena or non-nuclear, man-made phenomena. This document does not address requirements concerning acquisition, processing and analysis of radionuclide data, but includes requirements for the dissemination of radionuclide data and products.

LWR physics in SKODA Works

International Nuclear Information System (INIS)

Zbytovsky, A.; Lehmann, M.; Vyskocil, V.; Vacek, J.; Krysl, V.

1980-01-01

Computation of nuclear power reactors of the WWER-1000 type is described as are computer programs used by Skoda Works for the solution of neutron problems. The programs are analyzed for applicability in the unified program system of the CMEA countries which will be used in the preparation of safety reports, the evaluation of safety hazards, the design of fuel charges, economical studies etc. A detailed description is also presented of multigroup transport calculations and of the preparation of input data for macrocalculations of the heterogeneous lattices of LWR's. (author)

Westinghouse Hanford Company (WHC) standards/requirements identification document (S/RID)

International Nuclear Information System (INIS)

Bennett, G.L.

1996-01-01

This Standards/Requirements Identification Document (S/RID) set forth the Environmental Safety and Health (ES ampersand amp;H) standards/requirements for Westinghouse Hanford Company Level Programs, where implementation and compliance is the responsibility of these organizations. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment

Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

Energy Technology Data Exchange (ETDEWEB)

Kristine Barrett; Shannon Bragg-Sitton

2012-09-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

International Nuclear Information System (INIS)

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

2012-01-01

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

Energy Technology Data Exchange (ETDEWEB)

Permana, Sidik; Suzuki, Mitsutoshi; Su' ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

2012-06-06

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

Project W-236A, work plan for preparation of a design requirements document

International Nuclear Information System (INIS)

Groth, B.D.

1995-01-01

This work plan outlines the tasks necessary, and defines the organizational responsibilities for preparing a Design Requirements Document (DRD) for project W-236A, Multi-Function Waste Tank Facility (MWTF). A DRD is a Systems Engineering document which bounds, at a high level, the requirements of a discrete system element of the Tank Waste Remediation System (TWRS) Program. This system element is usually assigned to a specific project, in this case the MWTF. The DRD is the document that connects the TWRS program requirements with the highest level projects requirements and provides the project's link to the overall TWRS mission. The MWTF DRD effort is somewhat unique in that the project is already in detailed design, whereas a DRO is normally prepared prior to preliminary design. The MWTF design effort was initiated with a Functional Design Criteria (FDC) and a Supplemental Design Requirements Document (SDRD) bounding the high level requirements. Another unique aspect of this effort is that some of the TWRS program requirements are still in development. Because of these unique aspects of the MWTF DRD development, the MWTF will be developed from existing TWRS Program requirements and project specific requirements contained in the FDC and SDRD. The following list describes the objectives of the MWTF DRD: determine the primary functions of the tanks through a functional decomposition of the TWRS Program high level functions; allocate the primary functions to a sub-system architecture for the tanks; define the fundamental design features in terms of performance requirements for the system and subsystems; identify system interfaces and design constraints; and document the results in a DRD

CH2M Hill Hanford Group, Inc., Standards and Requirements Identification Document (SRID) Requirements Management System and Requirements Specification

International Nuclear Information System (INIS)

JOHNSON, A.L.

2000-01-01

The current Tank Farm Contractor (TFC) for the U. S. Department of Energy, Office of River Protection (ORP), River Protection Project (RPP), CH2M Hill Hanford Group, Inc. (CHG), will use a computer based requirements management system. The system will serve as a tool to assist in identifying, capturing, and maintaining the Standards/Requirements Identification Document (S/RID) requirements and links to implementing procedures and other documents. By managing requirements as one integrated set, CHG will be able to carry out its mission more efficiently and effectively. CHG has chosen the Dynamic Object Oriented Requirements System (DOORS(trademark)) as the preferred computer based requirements management system. Accordingly, the S/RID program will use DOORS(trademark). DOORS(trademark) will replace the Environmental Requirements Management Interface (ERMI) system as the tool for S/RID data management. The DOORS(trademark) S/RID test project currently resides on the DOORSTM test server. The S/RID project will be migrated to the DOORS(trademark) production server. After the migration the S/RID project will be considered a production project and will no longer reside on the test server

Plutonium Finishing Plant (PFP) Standards/Requirements Identification Document (S/RID)

Energy Technology Data Exchange (ETDEWEB)

Maddox, B.S.

1996-01-01

This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ESH) standards/requirements for the Plutonium Finishing Plant (PFP). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

Plutonium Finishing Plant (PFP) Standards/Requirements Identification Document (S/RID)

International Nuclear Information System (INIS)

Maddox, B.S.

1996-01-01

This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ESH) standards/requirements for the Plutonium Finishing Plant (PFP). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment

17 CFR 4.31 - Required delivery of Disclosure Document to prospective clients.

Science.gov (United States)

2010-04-01

... Disclosure Document to prospective clients. 4.31 Section 4.31 Commodity and Securities Exchanges COMMODITY... Advisors § 4.31 Required delivery of Disclosure Document to prospective clients. (a) Each commodity trading... prospective client a Disclosure Document containing the information set forth in §§ 4.34 and 4.35 for the...

International collaboration for development of accident-resistant LWR fuel. International Collaboration for Development of Accident Resistant Light Water Reactor Fuel

International Nuclear Information System (INIS)

Sowder, Andrew

2013-01-01

Following the March 2011 multi-unit accident at the Fukushima Daiichi plant, there has been increased interest in the development of breakthrough nuclear fuel designs that can reduce or eliminate many of the outcomes of a severe accident at a light water reactor (LWR) due to loss of core cooling following an extended station blackout or other initiating event. With this interest and attention comes a unique opportunity for the nuclear industry to fundamentally change the nature and impact of severe accidents. Clearly, this is no small feat. The challenges are many and the technical barriers are high. Early estimates for moving maturing R and D concepts to the threshold of commercialisation exceed one billion USD. Given the anticipated effort and resources required, no single entity or group can succeed alone. Accordingly, the Electric Power Research Institute (EPRI) sees the need for and promise of cooperation among many stakeholders on an international scale to bring about what could be transformation in LWR fuel performance and robustness. An important initial task in any R and D programme is to define the goals and metrics for measuring success. As starting points for accident-tolerant fuel development, the extension of core coolability under loss of coolant conditions and the elimination or reduction of hydrogen generation are widely recognised R and D endpoints for deployment. Furthermore, any new LWR fuel technology will, at a minimum, need to (1) be compatible with the safe, economic operation of existing plants and (2) maintain acceptable or improve nuclear fuel performance under normal operating conditions. While the primary focus of R and D to date has been on cladding and fuel improvements, there are a number of other potential paths to improve outcomes following a severe accident at an LWR that include modifications to other fuel hardware and core internals to fully address core coolability, criticality, and hydrogen generation concerns. The US

Design requirements document for project W-520, immobilized low-activity waste disposal

International Nuclear Information System (INIS)

Ashworth, S.C.

1998-01-01

This design requirements document (DRD) identifies the functions that must be performed to accept, handle, and dispose of the immobilized low-activity waste (ILAW) produced by the Tank Waste Remediation System (TWRS) private treatment contractors and close the facility. It identifies the requirements that are associated with those functions and that must be met. The functional and performance requirements in this document provide the basis for the conceptual design of the Tank Waste Remediation System Immobilized Low-Activity Waste disposal facility project (W-520) and provides traceability from the program-level requirements to the project design activity

Design requirements document for project W-520, immobilized low-activity waste disposal

Energy Technology Data Exchange (ETDEWEB)

Ashworth, S.C.

1998-08-06

This design requirements document (DRD) identifies the functions that must be performed to accept, handle, and dispose of the immobilized low-activity waste (ILAW) produced by the Tank Waste Remediation System (TWRS) private treatment contractors and close the facility. It identifies the requirements that are associated with those functions and that must be met. The functional and performance requirements in this document provide the basis for the conceptual design of the Tank Waste Remediation System Immobilized Low-Activity Waste disposal facility project (W-520) and provides traceability from the program-level requirements to the project design activity.

Environmental restoration remedial action quality assurance requirements document

International Nuclear Information System (INIS)

Cote, R.F.

1991-01-01

The environmental Restoration Remedial Action Quality Assurance Requirements Document (DOE/RL 90-28) defines the quality assurance program requirements for the US Department of Energy-Richland Field Office Environmental Restoration Remedial Action Program at the Hanford Site, Richland, Washington. This paper describes the objectives outlined in DOE/RL 90-28. The Environmental Restoration Remedial Action Program implements significant commitments made by the US Department of Energy in the Hanford Federal Facility Agreement and Consent Order entered into with the Washington State Department of Ecology and the US Environmental Protection Agency

Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-09-15

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U₃Si₂) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U₃Si₂ as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U₃Si₂ fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U₃Si₂. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

A study for small-medium LWR development of JAPC

International Nuclear Information System (INIS)

Okazaki, Toshihiko; Hida, Takahiko; Hoshi, Takashi; Kawahara, Hiroto; Tominaga, Kenji; Asano, Hiromitsu

2011-01-01

LWR (Light Water Reactor) power stations have accumulated many experiences of design, construction and operation. In addition, large-sized reactors have an advantage of economy of scale and 1,000 MWe LWR has therefore become the mainstream reactor in Japan. Meanwhile, introduction of the medium and small-sized LWRs (SMRs) has also been under review in Japan in order to respond to stagnant growth in electricity demand and electricity market liberalization or for investment risk mitigation; however, it has not been realized due to the economic disadvantage of scale. Therefore, JAPC has been developing the concept of SMR (300 MWe - 600 MWe) which is competitive to the large-sized LWR cooperating with Japanese plant makers (Hitachi, Toshiba Corporation and Mitsubishi Heavy Industries), assessing the possibility of realization of SMRs as one of the electric power sources in the future. As the result of the JAPC's study, we developed SMR concepts whose cost and safety are almost equal to large-sized LWR and confirmed technical feasibility of the concept in order to start developing basic design. In this paper, the outline of the SMR concepts and the current development status are presented. Concepts have been developed for two types of SMRs (i.e. BWR and PWR). As for the BWR type, reactor system is simplified by adopting natural circulation core method and CRD falling under gravity in order to downsize the reactor containments. As for the PWR type, the risk of LOCA occurrence is eliminated by unifying the primary system (e.g. incorporating steam generator into reactor). Furthermore, the primary system is simplified by adopting natural circulation core method in operation and containment vessel also become compact for the PWR. As for JAPC's further development of SMRs, key elements of SMR concepts are studied. In addition, the environment surrounding the SMRs has changed in recent years and the one with capacity exceeding 300-600 MWe class or small-sized reactor with

Safety aspects and operating experience of LWR plants in Japan

International Nuclear Information System (INIS)

Aoki, S.; Yoshioka, T.; Toyota, M.; Hinoki, M.

1977-01-01

To develop nuclear power generation for the future, it is necessary to put further emphasis on safety assurance and to endeavour to devise measures to improve plant availability, based on the careful analysis of causes that reduce plant availability. The paper discusses the results of studies on the following items from such viewpoints: (1) Safety and operating experience of LWR nuclear power plants in Japan: operating experience with LWRs; improvements in LWR design during the past ten years; analysis of the factors affecting plant availability; (2) Assurance of safety and measures to increase availability: measures for safety and environmental protection; measures to reduce radiation exposure of employees; appropriateness of maintenance and inspection work; measures to increase plant availability; measures to improve reliability of equipment and components; (3) Future technical problems. (author)

LWR aerosol containment experiments (LACE) program and initial test results

International Nuclear Information System (INIS)

Muhlestein, L.D.; Hilliard, R.K.; Bloom, G.R.; McCormack, J.D.; Rahn, F.J.

1985-01-01

The LWR aerosol containment experiments (LACE) program is described. The LACE program is being performed at the Hanford Engineer Development Laboratory (operated by Westinghouse Hanford Company) and the initial tests are sponsored by EPRI. The objectives of the LACE program are: to demonstrate, at large-scale, inherent radioactive aerosol retention behavior for postulated high consequence LWR accident situations; and to provide a data base to be used for aerosol behavior . Test results from the first phase of the LACE program are presented and discussed. Three large-scale scoping tests, simulating a containment bypass accident sequence, demonstrated the extent of agglomeration and deposition of aerosols occurring in the pipe pathway and vented auxiliary building under realistic accident conditions. Parameters varied during the scoping tests were aerosol type and steam condensation

50 CFR 23.19 - What CITES documents are required to export Appendix-I plants?

Science.gov (United States)

2010-10-01

... 50 Wildlife and Fisheries 6 2010-10-01 2010-10-01 false What CITES documents are required to... ENDANGERED SPECIES OF WILD FAUNA AND FLORA (CITES) Prohibitions, Exemptions, and Requirements § 23.19 What CITES documents are required to export Appendix-I plants? Answer the questions in the following decision...

Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

International Nuclear Information System (INIS)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

1986-01-01

The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) and ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models

Safety-related LWR research. Annual report 1993

International Nuclear Information System (INIS)

Hueper, R.

1994-06-01

The reactor safety R and D work of the Karlsruhe Nuclear Research Centre (KfK) has been part of the Nuclear Safety Research Project (PSF) since 1990. The present annual report 1993 summarizes the results on LWR safety. The research tasks are coordinated in agreement with internal and external working groups. The contributions to this report correspond to the status at the end of 1993. (orig./HP) [de

Economic analyses of LWR fuel cycles

International Nuclear Information System (INIS)

Field, F.R.

1977-05-01

An economic comparison was made of three options for handling irradiated light-water reactor (LWR) fuel. These options are reprocessing of spent reactor fuel and subsequent recycle of both uranium and plutonium, reprocessing and recycle of uranium only, and direct terminal storage of spent fuel not reprocessed. The comparison was based on a peak-installed nuclear capacity of 507 GWe by CY 2000 and retirement of reactors after 30 years of service. Results of the study indicate that: Through the year 2000, recycle of uranium and plutonium in LWRs saves about $12 billion (FY 1977 dollars) compared with the throwaway cycle, but this amounts to only about 1.3% of the total cost of generating electricity by nuclear power. If deferred costs are included for fuel that has been discharged from reactors but not reprocessed, the economic advantage increases to $17.7 billion. Recycle of uranium only (storage of plutonium) is approximately $7 billion more expensive than the throwaway fuel cycle and is, therefore, not considered an economically viable option. The throwaway fuel cycle ultimately requires >40% more uranium resources (U 3 O 8 ) than does reprocessing spent fuel where both uranium and plutonium are recycled

System requirements and design description for the document basis database interface (DocBasis)

International Nuclear Information System (INIS)

Lehman, W.J.

1997-01-01

This document describes system requirements and the design description for the Document Basis Database Interface (DocBasis). The DocBasis application is used to manage procedures used within the tank farms. The application maintains information in a small database to track the document basis for a procedure, as well as the current version/modification level and the basis for the procedure. The basis for each procedure is substantiated by Administrative, Technical, Procedural, and Regulatory requirements. The DocBasis user interface was developed by Science Applications International Corporation (SAIC)

EUV lithography for 22nm half pitch and beyond: exploring resolution, LWR, and sensitivity tradeoffs

Science.gov (United States)

Putna, E. Steve; Younkin, Todd R.; Leeson, Michael; Caudillo, Roman; Bacuita, Terence; Shah, Uday; Chandhok, Manish

2011-04-01

The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 22nm half pitch node and beyond. According to recent assessments made at the 2010 EUVL Symposium, the readiness of EUV materials remains one of the top risk items for EUV adoption. The main development issue regarding EUV resists has been how to simultaneously achieve high resolution, high sensitivity, and low line width roughness (LWR). This paper describes our strategy, the current status of EUV materials, and the integrated post-development LWR reduction efforts made at Intel Corporation. Data collected utilizing Intel's Micro- Exposure Tool (MET) is presented in order to examine the feasibility of establishing a resist process that simultaneously exhibits <=22nm half-pitch (HP) L/S resolution at <=11.3mJ/cm2 with <=3nm LWR.

A Framework for Requirement Elicitation, Analysis, Documentation and Prioritisation under Uncertainty

NARCIS (Netherlands)

Rajabali Nejad, Mohammadreza; Mladenov, V.

2015-01-01

This paper offers a pluralistic framework for coping with requirements in the early phases of design where there is lack of knowledge about a system, its architect and functions. The framework is used to elicit, analyze, document and prioritize the requirements. It embeds probabilistic approach and

Tools for LWR spent fuel characterization: Assembly classes and fuel designs

International Nuclear Information System (INIS)

Moore, R.S.; Notz, K.J.

1991-01-01

The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs

Fission product release from high gap-inventory LWR fuel under LOCA conditions

International Nuclear Information System (INIS)

Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

1980-01-01

Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Onoue, Masaaki; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

2000-11-01

A set of ORIGEN2 libraries for LWR MOX fuels was developed based on JENDL-3.2. The libraries were compiled with SWAT using the specification of MOX fuels that will be used in nuclear power reactors in Japan. The verification of the libraries were performed by the analyses of post irradiation examinations for the fuels from European PWR. By the analysis of PIE data from PWR in United States, the comparison was made between calculation and experimental results in the case of that parameters for making the libraries are different from irradiation conditions. These new libraries for LWR MOX fuels are packaged in ORLIBJ32, the libraries released in 1999. (author)

Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations

Energy Technology Data Exchange (ETDEWEB)

Brown, S. J. [ed.

1980-04-01

The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics listed are based upon the experience of ODAI and also based upon a sampling of over 100 engineers/scientists in the LWR industry. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation.

Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations

International Nuclear Information System (INIS)

Brown, S.J.

1980-04-01

The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics listed are based upon the experience of ODAI and also based upon a sampling of over 100 engineers/scientists in the LWR industry. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation

Workshop on initiation of stress corrosion cracking under LWR conditions: Proceedings

International Nuclear Information System (INIS)

Nelson, J.L.; Cubicciotti, D.; Licina, G.J.

1988-05-01

A workshop titled ''Initiation of Stress Corrosion Cracking under LWR Conditions'' was held in Palo Alto, California on November 13, 1986, hosted by the Electric Power Research Institute. Participants were experts on the topic from nuclear steam supply and component manufacturers, public and private research laboratories, and university environments. Presentations included discussions on the definition of crack initiation, the effects of environmental and electrochemical variables on cracking susceptibility, and detection methods for the determination of crack initiation events and measurement of critical environmental and stress parameters. Examination of the questions related to crack initiation and its relative importance to the overall question of cracking of LWR materials from these perspectives provided inputs to EPRI project managers on the future direction of research efforts designed to prevent and control cracking. Thirteen reports have been cataloged separately

An overview of advanced high-strength nickel-base alloys for LWR applications

International Nuclear Information System (INIS)

Prybylowski, J.; Ballinger, R.G.

1989-01-01

This paper reviews our current understanding of the behavior of high strength nickel base alloys used in light water reactor (LWR) applications. Emphasis is placed on understanding the fundamental mechanisms controlling crack propagation in these environments. To provide a foundation for this survey, general mechanisms of stress corrosion cracking and hydrogen embrittlement are first reviewed. The behavior of high strength nickel base alloys in LWR environments, as well as in other relevant environments is then reviewed. Suggested mechanisms of crack propagation are discussed. Alternate alloys and microstructural modifications that may result in improved behavior are presented. It is now clear that, at temperatures near 100C, alloy X-750, the predominant high strength nickel base alloy used today in LWR applications, is susceptible to hydrogen embrittlement. A review of published data from hydrogen embrittlement studies of nickel base superalloys during electrolytic charging and in hydrogen sulfide/brine solutions suggests that other nickel base superalloys are available possessing resistance to hydrogen embrittlement superior to that of alloy X-750. Available results of tests in gaseous hydrogen suggest that reduced grain boundary precipitation and a fine distribution of intragranular precipitates that act as irreversible hydrogen traps is the optimum microstructure for hydrogen embrittlement resistance. 42 refs., 2 figs., 5 tabs

Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

Energy Technology Data Exchange (ETDEWEB)

Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

2004-07-01

For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

International Nuclear Information System (INIS)

Purcell, P.C.; Dallongeville, M.

2004-01-01

For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

Supplemental design requirements document solid waste operations complex

International Nuclear Information System (INIS)

Ocampo, V.P.; Boothe, G.F.; Broz, D.R.; Eaton, H.E.; Greager, T.M.; Huckfeldt, R.A.; Kooiker, S.L.; Lamberd, D.L.; Lang, L.L.; Myers, J.B.

1994-11-01

This document provides additional and supplemental information to the WHC-SD-W112-FDC-001, WHC-SD-W113-FDC-001, and WHC-SD-W100-FDC-001. It provides additional requirements for the design and summarizes Westinghouse Hanford Company key design guidance and establishes the technical baseline agreements to be used for definitive design common to the Solid Waste Operations Complex (SWOC) Facilities (Project W-112, Project W-113, and WRAP 2A)

Waste encapsulation storage facility (WESF) standards/requirements identification document (S/RIDS)

Energy Technology Data Exchange (ETDEWEB)

Maddox, B.S., Westinghouse Hanford

1996-07-29

This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ES{ampersand}H) standards/requirements for the Waste Encapsulation Storage Facility (WESF). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

Program prioritization system user requirements document for Gas Cooled Reactor Associates

International Nuclear Information System (INIS)

1981-01-01

Efficient management of the national HTGR program requires the establishment of an information system that will facilitate a more rational allocation of resources and task prioritization consistent with program policies. The system described in this document provides a data analysis mechanism for processing top level summary status and planning information in a rapid, timely and selective manner. Data produced by the system can be used by management to provide a rational basis for prioritizing tasks, evaluating program changes and program planning regarding costs, schedules and overall program development logic. The purpose of this document is to delineate the program prioritization system (PPS) requirements for use as a guide to acquiring and implementing the system

Nuclear fuel requirements in Egypt up to the year 2000

International Nuclear Information System (INIS)

Hammad, F.H.; Ghoneim, M.M.; Abou-Zahra, A.A.

1976-01-01

The nuclear fuel demands (as U 3 O 8 ) in Egypt for the nuclear power and nuclear desalination programmes up to the year 2000 were calculated. The recent nuclear power forecasts of the International Atomic Energy Agency (IAEA) and the General Egyptian Electricity Coroporation (GEEC) were used as bases for calculation. The requirements were calculated for light and heavy water reactor (LWR and HWR) strategies. The cumulative uranium requirements for power and desalination programmes were found to be in the range of 19000-25000 tons in case of HWR strategy and 30000-400000 tons in case of LWR strategy. In the year 2000 alone the demands are expected to be 2000-2500 tons in case of HWR strategy and 3200-4000 tons in case of LWR strategy. Intensive uranium exploration coupled with a suitable reactor policy is essential to ensure the availability of these demands and for developing a local nuclear fuel technology

DOE Integrated Safeguards and Security (DISS) historical document archival and retrieval analysis, requirements and recommendations

Energy Technology Data Exchange (ETDEWEB)

Guyer, H.B.; McChesney, C.A.

1994-10-07

The overall primary Objective of HDAR is to create a repository of historical personnel security documents and provide the functionality needed for archival and retrieval use by other software modules and application users of the DISS/ET system. The software product to be produced from this specification is the Historical Document Archival and Retrieval Subsystem The product will provide the functionality to capture, retrieve and manage documents currently contained in the personnel security folders in DOE Operations Offices vaults at various locations across the United States. The long-term plan for DISS/ET includes the requirement to allow for capture and storage of arbitrary, currently undefined, clearance-related documents that fall outside the scope of the ``cradle-to-grave`` electronic processing provided by DISS/ET. However, this requirement is not within the scope of the requirements specified in this document.

Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

Energy Technology Data Exchange (ETDEWEB)

Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

2009-07-01

The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

Review and comparison of WWER and LWR Codes and Standards

International Nuclear Information System (INIS)

Buckthorpe, D.; Tashkinov, A.; Brynda, J.; Davies, L.M.; Cueto-Felgeueroso, C.; Detroux, P.; Bieniussa, K.; Guinovart, J.

2003-01-01

The results of work on a collaborative project on comparison of Codes and Standards used for safety related components of the WWER and LWR type reactors is presented. This work was performed on behalf of the European Commission, Working Group Codes and Standards and considers areas such as rules, criteria and provisions, failure mechanisms , derivation and understanding behind the fatigue curves, piping, materials and aging, manufacturing and ISI. WWERs are essentially designed and constructed using the Russian PNAE Code together with special provisions in a few countries (e.g. Czech Republic) from national standards. The LWR Codes have a strong dependence on the ASME Code. Also within Western Europe other codes are used including RCC-M, KTA and British Standards. A comparison of procedures used in all these codes and standards have been made to investigate the potential for equivalencies between the codes and any grounds for future cooperation between eastern and western experts in this field. (author)

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID)

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 3) presents the standards and requirements for the following sections: Safeguards and Security, Engineering Design, and Maintenance.

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID)

International Nuclear Information System (INIS)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 3) presents the standards and requirements for the following sections: Safeguards and Security, Engineering Design, and Maintenance

Light water reactors development in Japan. (1) Introduction of LWR technology (PWR)

International Nuclear Information System (INIS)

Yamada, Ichita; Suzuki, Shigemitsu

2008-01-01

Evolutionary progress of the LWR plants in the last half-century was reviewed in series. Introduction of LWR technology (PWR) in Japan was reviewed in this article. Kansai Electric Power imported the Mihama-1 - a 340 MWe PWR built by Westinghouse Corp. It began operating in 1970 to supply power to the World Exposition (EXPO70). There followed a period in which designs was purchased from US vendors and they were constructed with the co-operation of Mitsubishi Heavy Industry, who would then receive a license to build similar plants in Japan and develop the capacity to design and construct PWRs by itself. Progress of designs, fabrications, project management and construction of PWRs were reviewed from technology transfer to its autonomy age. (T. Tanaka)

Waste Receiving and Packaging, Module 2A, Supplemental Design Requirements Document

International Nuclear Information System (INIS)

Lamberd, D.L.; Boothe, G.F.; Hinkle, A.L.; Horgos, R.M.; LeClair, M.D.; Nash, C.R.; Ocampo, V.P.; Pauly, T.R.; Stroup, J.L.; Weingardt, K.M.

1994-01-01

The Supplemental Design Requirements Document (SDRD) is used to communicate plant design information from Westinghouse Hanford Company (WHC) to the US Department of Energy (DOE) and the cognizant Architect Engineer (A/E). Information in the SDRD serves two purposes: to convey design requirements that are too detailed for inclusion in a Functional Design Criteria (FDC) report; and to serve as a means of change control for design commitments in the Conceptual Design Report. The mission of WRAP 2A on the Hanford site is the treatment of contact handled low level mixed waste (MW) for final disposal. The overall systems engineering steps used to reach construction and operation of WRAP 2A are depicted in Figure 1. The WRAP 2A SDRD focuses on the requirements to address the functional analysis provided in Figure 1. This information is provided in sections 2 through 5 of this SDRD. The mission analysis and functional analysis are to be provided in a separate supporting document. The organization of sections 2 through 5 corresponds to the requirements identified in the WRAP 2A functional analysis

45 CFR 400.43 - Requirements for documentation of refugee status.

Science.gov (United States)

2010-10-01

... 45 Public Welfare 2 2010-10-01 2010-10-01 false Requirements for documentation of refugee status. 400.43 Section 400.43 Public Welfare Regulations Relating to Public Welfare OFFICE OF REFUGEE RESETTLEMENT, ADMINISTRATION FOR CHILDREN AND FAMILIES, DEPARTMENT OF HEALTH AND HUMAN SERVICES REFUGEE...

Safety criteria related to microheterogeneities in LWR mixed oxide fuels

International Nuclear Information System (INIS)

Renard, A.; Mostin, N.

1978-01-01

The main safety aspets of PuO 2 microheterogeneities in the pellets of LWR mixed oxide fuels are reviewed. Points of interest are studied, especially the transient behaviour in accidental conditions and criteria are deduced for use in the specification and quality control of the fabricated product. (author)

Qualification of ARROTTA code for LWR accident analysis

International Nuclear Information System (INIS)

Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.

2004-01-01

This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)

Effects of LWR coolant environments on fatigue lives of austenitic stainless steels

International Nuclear Information System (INIS)

Chopra, O.K.; Gavenda, D.J.

1997-01-01

The ASME Boiler and Pressure Vessel Code fatigue design curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue life of pressure vessel and piping materials in light water reactor (LWR) environments. Fatigue tests have been conducted on Types 304 and 316NG stainless steel in air and LWR environments to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on fatigue lives of these steels. The results confirm the significant decrease in fatigue life in water. The environmentally assisted decrease in fatigue life depends both on strain rate and DO content in water. A decrease in strain rate from 0.4 to 0.004%/s decreases fatigue life by a factor of ∼ 8. However, unlike carbon and low-alloy steels, environmental effects are more pronounced in low-DO than in high-DO water. At ∼ 0.004%/s strain rate, reduction in fatigue life in water containing <10 ppb D is greater by a factor of ∼ 2 than in water containing ≥ 200 ppb DO. Experimental results have been compared with estimates of fatigue life based on the statistical model. The formation and growth of fatigue cracks in austenitic stainless steels in air and LWR environments are discussed

Hanford analytical services quality assurance requirements documents. Volume 1: Administrative Requirements

International Nuclear Information System (INIS)

Hyatt, J.E.

1997-01-01

Hanford Analytical Services Quality Assurance Requirements Document (HASQARD) is issued by the Analytical Services, Program of the Waste Management Division, US Department of Energy (US DOE), Richland Operations Office (DOE-RL). The HASQARD establishes quality requirements in response to DOE Order 5700.6C (DOE 1991b). The HASQARD is designed to meet the needs of DOE-RL for maintaining a consistent level of quality for sampling and field and laboratory analytical services provided by contractor and commercial field and laboratory analytical operations. The HASQARD serves as the quality basis for all sampling and field/laboratory analytical services provided to DOE-RL through the Analytical Services Program of the Waste Management Division in support of Hanford Site environmental cleanup efforts. This includes work performed by contractor and commercial laboratories and covers radiological and nonradiological analyses. The HASQARD applies to field sampling, field analysis, and research and development activities that support work conducted under the Hanford Federal Facility Agreement and Consent Order Tri-Party Agreement and regulatory permit applications and applicable permit requirements described in subsections of this volume. The HASQARD applies to work done to support process chemistry analysis (e.g., ongoing site waste treatment and characterization operations) and research and development projects related to Hanford Site environmental cleanup activities. This ensures a uniform quality umbrella to analytical site activities predicated on the concepts contained in the HASQARD. Using HASQARD will ensure data of known quality and technical defensibility of the methods used to obtain that data. The HASQARD is made up of four volumes: Volume 1, Administrative Requirements; Volume 2, Sampling Technical Requirements; Volume 3, Field Analytical Technical Requirements; and Volume 4, Laboratory Technical Requirements. Volume 1 describes the administrative requirements

Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

International Nuclear Information System (INIS)

Miller, W.J.

1980-05-01

This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan

Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

Energy Technology Data Exchange (ETDEWEB)

Miller, W.J.

1980-05-01

This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan.

Analysis of compatibility of current Czech initial documentation in the area of technical assurance of nuclear safety with the requirements of the EUR document

International Nuclear Information System (INIS)

Zdebor, J.; Zdebor, R.; Kratochvil, L.

2001-11-01

The publication is structured as follows: Description of existing documentation. General requirements, goals, principles and design principles: Documents being compared; Method of comparison; Results and partial evaluation of comparison of requirements between EUR and Czech regulations (basic goals and safety philosophy; quantitative safety objectives; basic design requirements; extended design requirements; external and internal threats; technical requirements; site conditions); Summary of the comparison of safety requirements. Comparison of requirements for the systems: Requirements for the nuclear reactor unit systems; Barrier systems (fuel system; reactor cooling system; containment system); Remaining systems (control systems; protection systems; coolant makeup and purification system; residual heat removal system; emergency cooling system; power systems); Common technical requirements for systems (technical requirements for systems; internal and external events). (P.A.)

50 CFR 23.25 - What additional information is required on a non-Party CITES document?

Science.gov (United States)

2010-10-01

... a non-Party CITES document? 23.25 Section 23.25 Wildlife and Fisheries UNITED STATES FISH AND... IN ENDANGERED SPECIES OF WILD FAUNA AND FLORA (CITES) Prohibitions, Exemptions, and Requirements § 23.25 What additional information is required on a non-Party CITES document? (a) Purpose. Under Article...

Standard casks for the transport of LWR spent fuel. Storage/transport casks for long cooled spent fuel

International Nuclear Information System (INIS)

Blum, P.; Sert, G.; Gagnon, R.

1983-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks are presently used for European and intercontinental transports and manufactured under TRANSNUCLEAIRE supervision in different countries. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardization faciliting fabrication, operation and spare part supply. Recently, TRANSNUCLEAIRE also developed a new generation of casks for the dry storage and occasional transport of LWR spent fuel which has been cooled for 5 years or 7 years in case of consolidated fuel rods. These casks have an optimum payload which takes into account the shielding requirements and the weight limitations at most sites. This paper deals more particularly with the TN 24 model which exists in 4 versions among which one for 24 PWR 900 fuel assemblies and another one for the consolidated fuel rods from 48 of same fuel assemblies

The dupic fuel cycle synergism between LWR and HWR

International Nuclear Information System (INIS)

Lee, J.S.; Yang, M.S.; Park, H.S.; Lee, H.H.; Kim, K.P.; Sullivan, J.D.; Boczar, P.G.; Gadsby, R.D.

1999-01-01

The DUPIC fuel cycle can be developed as an alternative to the conventional spent fuel management options of direct disposal or plutonium recycle. Spent LWR fuel can be burned again in a HWR by direct refabrication into CANDU-compatible DUPIC fuel bundles. Such a linkage between LWR and HWR can result in a multitude of synergistic effects, ranging from savings of natural uranium to reductions in the amount of spent fuel to be buried in the earth, for a given amount of nuclear electricity generated. A special feature of the DUPIC fuel cycle is its compliance with the 'Spent Fuel Standard' criteria for diversion resistance, throughout the entire fuel cycle. The DUPIC cycle thus has a very high degree of proliferation resistance. The cost penalty due to this technical factor needs to be considered in balance with the overall benefits of the DUPIC fuel cycle. The DUPIC alternative may be able to make a significant contribution to reducing spent nuclear fuel burial in the geosphere, in a manner similar to the contribution of the nuclear energy alternative in reducing atmospheric pollution from fossil fuel combustion. (author)

48 CFR 752.7005 - Submission requirements for development experience documents.

Science.gov (United States)

2010-10-01

... (referenced in paragraph (a)(1) of this clause) in electronic format and hard copy (one copy) to U.S. Agency... final equivalent of the hard copy submitted. (iv) Acceptable software formats for electronic documents... paragraph (b)(1)(i) of this clause. (2) Format. (i) Descriptive information is required for all Contractor...

Modular approach to LWR in-core fuel management

International Nuclear Information System (INIS)

Urli, N.; Pevec, D.; Coffou, E.; Petrovic, B.

1980-01-01

The most important methods in the LWR in-core fuel management are reviewed. A modular approach and optimization by use of infinite multiplication factor and power form-factor are favoured. A computer program for rotation of fuel assemblies at reloads has been developed which improves further fuel economy and reliability of nuclear power plants. The program has been tested on the PWR core and showed to decrease the power form-factors and flatten the radial power distribution. (author)

Fiscal Year 2001 Tank Characterization Technical Sampling Basis and Waste Information Requirements Document

International Nuclear Information System (INIS)

ADAMS, M.R.

2000-01-01

The Fiscal Year 2001 Tank Characterization Technical Sampling Basis and Waste Information Requirements Document (TSB-WIRD) has the following purposes: (1) To identify and integrate sampling and analysis needs for fiscal year (FY) 2001 and beyond. (2) To describe the overall drivers that require characterization information and to document their source. (3) To describe the process for identifying, prioritizing, and weighting issues that require characterization information to resolve. (4) To define the method for determining sampling priorities and to present the sampling priorities on a tank-by-tank basis. (5) To define how the characterization program is going to satisfy the drivers, close issues, and report progress. (6)To describe deliverables and acceptance criteria for characterization deliverables

Experimental critical loadings and control rod worths in LWR-PROTEUS configurations compared with MCNPX results

International Nuclear Information System (INIS)

Plaschy, M.; Murphy, M.; Jatuff, F.; Seiler, R.; Chawla, R.

2006-01-01

The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of the PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor k eff (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)

Evaluation of LWR fuel rod behavior under operational transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Hiramoto, K.; Maru, A.

1984-01-01

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

Preliminary concepts for detecting national diversion of LWR spent fuel

International Nuclear Information System (INIS)

Sonnier, C.S.; Cravens, M.N.

1978-04-01

Preliminary concepts for detecting national diversion of LWR spent fuel during storage, handling and transportation are presented. Principal emphasis is placed on means to achieve timely detection by an international authority. This work was sponsored by the Department of Energy/Office of Safeguards and Security (DOE/OSS) as part of the overall Sandia Fixed Facility Physical Protection Program

LWR pressure vessel irradiation surveillance dosimetry. Quarterly progress report, July--September 1978

Energy Technology Data Exchange (ETDEWEB)

Guthrie, G L; McElroy, W N; Lippincott, E P; Gold, R

1978-12-01

Program objectives and progress to date by the national laboratories in LWR pressure vessel irradiation surveillance dosimetry are summarized. Participants in the program include: Rockwell International, Hanford Engineering Development Laboratory, National Bureau of Standards, and Oak Ridge National Laboratory.

Experience gained in the current LWR that influence the design and operation of the LWR advanced from the viewpoint of safety analysis

International Nuclear Information System (INIS)

Barrera, J.; Corisco, M.; Riverola, J.

2010-01-01

Since the construction of the first light water reactors (LWR) safety analysis has played a very important role in the operation and its evolution to come up with designs that are currently operating. With new tools available, this role will see increased allowing more efficient operation with security assessments in real time, and a more efficient designs both in terms of fuel efficiency and from the security of the plant during operation.

Behaviour of LWR core materials under accident conditions. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-12-01

At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs

Qualification of the neutronic evolution of LWR fuels in MELUSINE

International Nuclear Information System (INIS)

Beretz, D.; Garcin, J.; Ducros, G.; Vanhumbeeck, D.; Chaucheprat, P.

1984-09-01

MELUSINE, a swimming pool type reactor, in Grenoble, for research and technological irradiations is well fitted to the neutronic evolution qualification of the LWR fuel. Thus, with an adjustment of the lattice pitch, representative neutron spectrum locations are available. The re-leading management and the regulation mode flexibility of MELUSINE lead to reproductible neutronic parameters configurations without restricting the reactor to this purpose only. Under these conditions, simple calculations can be carried out for interpretation, without taking into account the whole core. An instrumentation by Self Power Neutron Detectors (collectrons) gives on-line information on the fluxes at the periphery of the device. When required by the neutronicians, experimental pins can be unloaded during the irradiation process and scanned on a gammametry bench immersed in the reactor-pool itself, before their isotopic composition analysis. Thus, within the framework of neutronic evolution qualification, are studied fuel pins for advanced assemblies for the light water reactors or their derivatives, with large advantages over irradiations in power reactors [fr

Fluidized-bed calcination of LWR fuel-reprocessing HLLW: requirements and potential for off-gas cleanup

International Nuclear Information System (INIS)

Schindler, R.E.

1979-01-01

Fluidized-bed solidification (calcination) was developed on a pilot scale for a variety of simulated LWR high-level liquid-waste (HLLW) and blended high-level and intermediate-level liquid-waste (ILLW) compositions. It has also been demonstrated with ICPP fuel-reprocessing waste since 1963 in the Waste Calcining Facility (WCF) at gross feed rates of 5 to 12 m 3 /day. A fluidized-bed calciner produces a relatively large volume of off-gas. A calciner solidifying 6 m 3 /day of liquid waste would generate about 13 standard m 3 /min of off-gas containing 10 to 20 g of entrained solids per standard m 3 of off-gas. Use of an off-gas system similar to that of the WCF could provide an overall process decontamination factor for particulates of about 2 x 10 10 . A potential advantage of fluidized-bed calcination over other solidification methods is the ability to control ruthenium volatilization from the calciner at less than 0.01% by calcining at 500 0 C or above. Use of an off-gas system similar to that of the WCF would provide an overall process decontamination factor for volatile ruthenium of greater than 1.6 x 10 7

Safety research for LWR type reactors

International Nuclear Information System (INIS)

1989-07-01

The current R and D activities are to be seen in connection with the LWR risk assessment studies. Two trends are emerging, of which the one concentrates more on BWR-specific problems, and the other on the efficiency or safety-related assessment of accident management activities. This annual report of 1988 reviews the progress of work done by the institutes and departments of the Karlsruhe Nuclear Research Center, (KfK), or on behalf of KfK by external institutions, in the field of safety research. The papers of this report present the state of work at the end of the year 1988. They are written in German, with an abstract in English. (orig./HP) [de

Proceedings of the IAEA specialists` meeting on cracking in LWR RPV head penetrations

Energy Technology Data Exchange (ETDEWEB)

Pugh, C.E.; Raney, S.J. [comps.] [Oak Ridge National Lab., TN (United States)

1996-07-01

This report contains 17 papers that were presented in four sessions at the IAEA Specialists` meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately.

Proceedings of the IAEA specialists' meeting on cracking in LWR RPV head penetrations

International Nuclear Information System (INIS)

Pugh, C.E.; Raney, S.J.

1996-07-01

This report contains 17 papers that were presented in four sessions at the IAEA Specialists' meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately

APEX nuclear fuel cycle for production of LWR fuel and elimination of radioactive waste

International Nuclear Information System (INIS)

Steinberg, M.; Powell, J.R.

1981-08-01

The development of a nuclear fission fuel cycle is proposed which eliminates all the radioactive fission product waste effluent and the need for geological-age high level waste storage and provides a long term supply of fissile fuel for an LWR power reactor economy. The fuel cycle consists of reprocessing LWR spent fuel (1 to 2 years old) to remove the stable nonradioactive (NRFP, e.g. lanthanides, etc.) and short-lived fission products SLFP e.g. half-lives of (1 to 2 years) and returning, in dilute form, the long-lived fission products, ((LLFPs, e.g. 30 y half-life Cs, Sr, and 10 y Kr, and 16 x 10 6 y I) and the transuranics (TUs, e.g. Pu, Am, Cm, and Np) to be refabricated into fresh fuel elements. Makeup fertile and fissile fuel are to be supplied through the use of a Spallator (linear accelerator spallation-target fuel-producer). The reprocessing of LWR fuel elements is to be performed by means of the Chelox process which consists of Airox treatment (air oxidation and hydrogen reduction) followed by chelation with an organic reagent (β-diketonate) and vapor distillation of the organometallic compounds for separation and partitioning of the fission products

50 CFR 23.53 - What are the requirements for obtaining a retrospective CITES document?

Science.gov (United States)

2010-10-01

... retrospective CITES document? 23.53 Section 23.53 Wildlife and Fisheries UNITED STATES FISH AND WILDLIFE SERVICE... ENDANGERED SPECIES OF WILD FAUNA AND FLORA (CITES) Application Procedures, Criteria, and Conditions § 23.53 What are the requirements for obtaining a retrospective CITES document? (a) Purpose. Retrospective...

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 4

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 4) presents the standards and requirements for the following sections: Radiation Protection and Operations.

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 4

International Nuclear Information System (INIS)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 4) presents the standards and requirements for the following sections: Radiation Protection and Operations

Effects of cooling time on a closed LWR fuel cycle

International Nuclear Information System (INIS)

Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

2012-01-01

In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov

2014-04-01

Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 5

International Nuclear Information System (INIS)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 5) outlines the standards and requirements for the Fire Protection and Packaging and Transportation sections

Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

International Nuclear Information System (INIS)

Whitesides, G.H.; Stephens, M.E.

1984-01-01

During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

Tank waste remediation system privatization infrastructure program requirements and document management process guide

International Nuclear Information System (INIS)

ROOT, R.W.

1999-01-01

This guide provides the Tank Waste Remediation System Privatization Infrastructure Program management with processes and requirements to appropriately control information and documents in accordance with the Tank Waste Remediation System Configuration Management Plan (Vann 1998b). This includes documents and information created by the program, as well as non-program generated materials submitted to the project. It provides appropriate approval/control, distribution and filing systems

Safety requirements applicable to the SMART design

International Nuclear Information System (INIS)

Seul, Kwang Won; Kim, Wee Kyong; Kim, Hho Jung

1999-01-01

The 330 MW thermal power of integral reactor, named SMART (System integrated Modular Advanced ReacTor), is under development at KAERI for seawater desalination application and electricity generation. The final product of nuclear desalination plant (NDP) is electricity and fresh water. Thus, in addition to the protection of the public around the plant facility from the possible release of radioactive materials, the fresh water should be prevented from radioactivity contamination. In this study, to ensure the safety of SMART reactor in the early stage of design development, the safety requirements applicable to the SMART design were investigated, based on the current regulatory requirements for the existing NPPs and the advanced light water reactor (LWR) designs. The interface requirements related to the desalination facility were also investigated, based on the recent IAEA research activities pertaining to the NDP. As a result, it was found that the current regulatory requirements and guidance for the existing NPPs and advanced LWR designs are applicable to the SMART design and its safety evaluation. However, the safety requirements related to the SMART-specific design and the desalination plant are needed to develop in the future to assure the safety of the SMART reactor

Functional Requirements Document for HALE UAS Operations in the NAS: Step 1. Version 3

Science.gov (United States)

2006-01-01

The purpose of this Functional Requirements Document (FRD) is to compile the functional requirements needed to achieve the Access 5 Vision of "operating High Altitude, Long Endurance (HALE) Unmanned Aircraft Systems (UAS) routinely, safely, and reliably in the national airspace system (NAS)" for Step 1. These functional requirements could support the development of a minimum set of policies, procedures and standards by the Federal Aviation Administration (FAA) and various standards organizations. It is envisioned that this comprehensive body of work will enable the FAA to establish and approve regulations to govern safe operation of UAS in the NAS on a routine or daily "file and fly" basis. The approach used to derive the functional requirements found within this FRD was to decompose the operational requirements and objectives identified within the Access 5 Concept of Operations (CONOPS) into the functions needed to routinely and safely operate a HALE UAS in the NAS. As a result, four major functional areas evolved to enable routine and safe UAS operations for an on-demand basis in the NAS. These four major functions are: Aviate, Navigate, Communicate, and Avoid Hazards. All of the functional requirements within this document can be directly traceable to one of these four major functions. Some functions, however, are traceable to several, or even all, of these four major functions. These cross-cutting functional requirements support the "Command / Control: function as well as the "Manage Contingencies" function. The requirements associated to these high-level functions and all of their supporting low-level functions are addressed in subsequent sections of this document.

NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

International Nuclear Information System (INIS)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO 2 fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO[sub 2] fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO{sub 2} fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

Standard casks for the transport of LWR spent fuel

International Nuclear Information System (INIS)

Blum, P.

1986-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufactured in different countries and are presently used for european and intercontinental transports. The main advantages of these casks are: large payload, moderate cost, reliability, standardisation facilitating fabrication, operation and spare part supply [fr

Investigation of valve failure problems in LWR power plants

International Nuclear Information System (INIS)

1980-04-01

An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems

European Utility Requirements (EUR) - the organisation and its products

International Nuclear Information System (INIS)

Ingemarsson, K.F.

2007-01-01

The idea of a common specification for the development and construction of new nuclear power plants appeared in Usa in the eighties in the framework of the ALWR program (Advanced Light Water Reactors). Several European electricity producers had participated in the writing of the EPRI -Utility Requirement Document (URD)- together with a group of American and Asian companies, but in the early nineties, they also agreed to produce a European document. The point was to write a more open specification (open to non-US designs) that would take into account specific European requirements, in nuclear safety in particular, while still keeping strong references to the EPRI URD. In late 1991, five of the major European electricity producers set up an organisation to develop the EUR document. Their primary objective was to produce a common set of requirements that could be endorsed by the major European electricity producers and that would provide clear guidance to the designers. The EUR document (EUR stands for European Utility Requirements) was born. Building new nuclear plants would require undisputable competitiveness vs. alternate production sources. For that only standardisation could bring an adequate answer in the coming years. Standardisation of the designs call for harmonization of the design rules, especially the ones related to nuclear safety. The EUR utilities support the initiatives that would pave the way to harmonization of the safety design rules at European level. They nevertheless think that it may be needed to go further and eventually consider an overall re-optimisation of the safety design rules. All these developments shall be organised at European level, keeping strong connections with the other methodological works undertaken outside Europe. The European utilities and the vendors have now an updated and well-tuned tool that allows them to develop, to assess and to order modern LWR designs well fitted to their actual needs. It has been used as the base

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 2

International Nuclear Information System (INIS)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Document (S/RID) is contained in multiple volumes. This document (Volume 2) presents the standards and requirements for the following sections: Quality Assurance, Training and Qualification, Emergency Planning and Preparedness, and Construction

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 2

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Document (S/RID) is contained in multiple volumes. This document (Volume 2) presents the standards and requirements for the following sections: Quality Assurance, Training and Qualification, Emergency Planning and Preparedness, and Construction.

Software requirements specification document for the AREST code development

International Nuclear Information System (INIS)

Engel, D.W.; McGrail, B.P.; Whitney, P.D.; Gray, W.J.; Williford, R.E.; White, M.D.; Eslinger, P.W.; Altenhofen, M.K.

1993-11-01

The Analysis of the Repository Source Term (AREST) computer code was selected in 1992 by the U.S. Department of Energy. The AREST code will be used to analyze the performance of an underground high level nuclear waste repository. The AREST code is being modified by the Pacific Northwest Laboratory (PNL) in order to evaluate the engineered barrier and waste package designs, model regulatory compliance, analyze sensitivities, and support total systems performance assessment modeling. The current version of the AREST code was developed to be a very useful tool for analyzing model uncertainties and sensitivities to input parameters. The code has also been used successfully in supplying source-terms that were used in a total systems performance assessment. The current version, however, has been found to be inadequate for the comparison and selection of a design for the waste package. This is due to the assumptions and simplifications made in the selection of the process and system models. Thus, the new version of the AREST code will be designed to focus on the details of the individual processes and implementation of more realistic models. This document describes the requirements of the new models that will be implemented. Included in this document is a section describing the near-field environmental conditions for this waste package modeling, description of the new process models that will be implemented, and a description of the computer requirements for the new version of the AREST code

Army Materiel Requirements Documents: Qualitative Analysis of Efficiency and Effectiveness

Science.gov (United States)

2013-06-01

focuses on the program’s time to execute their mission based off the MRDs/MCDs. We measure efficiency based on two BPP initiatives: (1) Build...definitions of each BPP initiative during our interviews. A poor rating indicates the requirement documents did not sufficiently support the program...in terms of efficiency and effectiveness according to BPP initiatives. For efficiency, we assign a qualitative measure based on SME responses across

Reduction of nuclear waste burden from LWR by deployment of the SCNES

International Nuclear Information System (INIS)

Arie, Kazuo; Watanabe, Junko; Mori, Kenji; Kubota, Kenichi; Kawashima, Masatoshi; Nakayama, Yoshiyuki; Nakazono, Ryuichi; Kuroda, Yuji; Fujiie, Yoichi

2009-01-01

Current efforts for enhancing capabilities for energy generation by LWR systems are efficient against the global warming crisis. In parallel to those movements, early realization of the SCNES concept can be the most viable solution to reduce nuclear waste burden produced by the current energy production system. (author)

Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-07-25

The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

International Nuclear Information System (INIS)

Heuser, Brent; Stubbins, James; Kozlowski, Tomasz; Uddin, Rizwan; Trinkle, Dallas; Downar, Thoms; Was, Gary; Ang, Yong; Phillpot, Simon; Sabharwall, Piyush

2017-01-01

The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

Environmental development plan. LWR commercial waste management

International Nuclear Information System (INIS)

1980-08-01

This Environmental Development Plan (EDP) identifies the planning and managerial requirements and schedules needed to evaluate and assess the environmental, health and safety (EH and S) aspects of the Commercial Waste Management Program (CWM). Environment is defined in its broadest sense to include environmental, health (occupational and public), safety, socioeconomic, legal and institutional aspects. This plan addresses certain present and potential Federal responsibilities for the storage, treatment, transfer and disposal of radioactive waste materials produced by the nuclear power industry. The handling and disposal of LWR spent fuel and processed high-level waste (in the event reprocessing occurs) are included in this plan. Defense waste management activities, which are addressed in detail in a separate EDP, are considered only to the extent that such activities are common to the commercial waste management program. This EDP addresses three principal elements associated with the disposal of radioactive waste materials from the commercial nuclear power industry, namely Terminal Isolation Research and Development, Spent Fuel Storage and Waste Treatment Technology. The major specific concerns and requirements addressed are assurance that (1) radioactivity will be contained during waste transport, interim storage or while the waste is considered as retrievable from a repository facility, (2) the interim storage facilities will adequately isolate the radioactive material from the biosphere, (3) the terminal isolation facility will isolate the wastes from the biosphere over a time period allowing the radioactivity to decay to innocuous levels, (4) the terminal isolation mode for the waste will abbreviate the need for surveillance and institutional control by future generations, and (5) the public will accept the basic waste management strategy and geographical sites when needed

Special equipment for the fabrication and quality control of nuclear fuel elements

International Nuclear Information System (INIS)

Guse, K.; Herbert, W.; Jaeger, K.

1989-01-01

For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG) [de

Proceedings of the 2007 LWR Fuel Performance Meeting / TopFuel 2007 'Zero by 2010'

International Nuclear Information System (INIS)

2007-01-01

ANS, ENS, AESJ and KNS are jointly organizing the 2007 International LWR Fuel Performance Meeting following the successful ENS TopFuel meeting held during 22-26 October, 2006 in Salamaca, Spain. Merging three premier nuclear fuel design and performance meetings: the ANS LWR Fuel Performance Meeting, the ENS TopFuel and Asian Water Reactor Fuel Performance Meeting (WRFPM) created this international meeting. The meeting will be held annually on a tri-annual rotational basis in USA, Asia, and Europe. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as performance experience in commercial and test reactors. The meeting excludes front end and back end fuel issues, however, it covers all front and/or back issues that impact fuel designs and performance

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector - 15271

International Nuclear Information System (INIS)

Saas, L.; Le Tellier, R.; Bajard, S.

2015-01-01

In this document, we present a simplified geometrical model (0D model) for both the in-core corium propagation transient and the characterization of the mode of corium transfer from the core to the vessel. A degraded core with a formed corium pool is used as an initial state. This initial state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate...). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4

High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety.

High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6

International Nuclear Information System (INIS)

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety

Investigation of valve failure problems in LWR power plants

Energy Technology Data Exchange (ETDEWEB)

None

1980-04-01

An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

Short Communication on "In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures"

Science.gov (United States)

Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.

2017-02-01

The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

LWR mox fuel experience in Belgium and France with special emphasis on results obtained in BR3

International Nuclear Information System (INIS)

Bairiot, H.; Haas, D.; Lippens, M.; Motte, F.; Lebastard, G.; Marin, J.F.

1986-09-01

The course of the paper reflects two main topics: LWR MOX fuel experience in Belgium and France, summarizing the fabrication techniques, the references, the underlying MOX fuel technology and the current R and D programs for expanding the data base; behaviour of MOX fuel rods irradiated under steady state and transient operating conditions, focusing on MOX fuel technology features acquired through the irradiations performed in the BR3 PWR, supplemented by tests in the BR2 MTR. This paper focuses on the thermomechanical behaviour of LWR MOX fuel rods, which is intimately related to the fabrication technique and vice-versa. 22 refs

42 CFR 495.336 - Health information technology planning advance planning document requirements (HIT PAPD).

Science.gov (United States)

2010-10-01

... 42 Public Health 5 2010-10-01 2010-10-01 false Health information technology planning advance... STANDARDS FOR THE ELECTRONIC HEALTH RECORD TECHNOLOGY INCENTIVE PROGRAM Requirements Specific to the Medicaid Program § 495.336 Health information technology planning advance planning document requirements...

23 CFR 420.111 - What are the documentation requirements for use of FHWA planning and research funds?

Science.gov (United States)

2010-04-01

... and Research Funds § 420.111 What are the documentation requirements for use of FHWA planning and... 23 Highways 1 2010-04-01 2010-04-01 false What are the documentation requirements for use of FHWA planning and research funds? 420.111 Section 420.111 Highways FEDERAL HIGHWAY ADMINISTRATION, DEPARTMENT OF...

LWR nuclear power plant component failures

International Nuclear Information System (INIS)

Schmidt, W.H.

1980-10-01

An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented

Step 1: Human System Interface (HSI) Functional Requirements Document (FRD). Version 2

Science.gov (United States)

2006-01-01

This Functional Requirements Document (FRD) establishes a minimum set of Human System Interface (HSI) functional requirements to achieve the Access 5 Vision of "operating High Altitude, Long Endurance (HALE) Unmanned Aircraft Systems (UAS) routinely, safely, and reliably in the National Airspace System (NAS)". Basically, it provides what functions are necessary to fly UAS in the NAS. The framework used to identify the appropriate functions was the "Aviate, Navigate, Communicate, and Avoid Hazards" structure identified in the Access 5 FRD. As a result, fifteen high-level functional requirements were developed. In addition, several of them have been decomposed into low-level functional requirements to provide more detail.

EUV lithography for 30nm half pitch and beyond: exploring resolution, sensitivity, and LWR tradeoffs

Science.gov (United States)

Putna, E. Steve; Younkin, Todd R.; Chandhok, Manish; Frasure, Kent

2009-03-01

The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 32nm half-pitch node and beyond. Readiness of EUV materials is currently one high risk area according to assessments made at the 2008 EUVL Symposium. The main development issue regarding EUV resist has been how to simultaneously achieve high sensitivity, high resolution, and low line width roughness (LWR). This paper describes the strategy and current status of EUV resist development at Intel Corporation. Data is presented utilizing Intel's Micro-Exposure Tool (MET) examining the feasibility of establishing a resist process that simultaneously exhibits <=30nm half-pitch (HP) L/S resolution at <=10mJ/cm2 with <=4nm LWR.

Hydrogen mixing study (HMS) in LWR type containments

International Nuclear Information System (INIS)

Travis, J.R.

1983-01-01

A numerical technique has been developed for calculating the full three-dimensional time-dependent Navier-Stokes equations with multiple speies transport. The method is a modified form of the Implicit Continuous-fluid Eulerian (ICE) technique to solve the governing equations for low Mach number flows where pressure waves and local variations in compression and expansion are not significant. Large density variations, due to thermal and species concentration gradients, are accounted for without the restrictions of the classical Boussinesq approximation. Calculations of the EPRI/HEDL standard problems verify the feasibility of using this finite-difference technique for analyzing hydrogen mixing within LWR containments

Standards/requirements identification documents (S/RIDS)

International Nuclear Information System (INIS)

Beckman, W.H.; Alhadeff, N.

1994-01-01

This paper describes the Fernald Environmental Restoration Management Corporation's (FERMCO) Standards/Requirement Identification Documents (S/RIDs) Development Program, the unique process used to implement it, and the status of the program. We will also discuss the lessons learned as the development program was implemented. The Department of Energy (DOE) established the Fernald site to produce uranium metals for the nation's defense programs in 1953. In 1989, DOE suspended production and, in 1991, the mission of the site was formally changed to one of environmental cleanup and restoration. The site was renamed the Fernald Environmental Management Project (FEMP) to reflect this change. From its inception until November 1992, the site was managed under a Management and Operating contract. As a result in the change in mission, DOE awarded an Environmental Restoration Management Contract (ERMC), focusing on restoration. FERMCO assumed management of the site December 1, 1992. The joint DOE/FERMCO mission is to protect human health and the environment through the safe, early, and least-cost final clean-up of the site in compliance with all applicable regulations and commitments while addressing stakeholder concerns. DOE has managed nuclear facilities primarily through its oversight of Management and Operating contractors. These contractors were responsible for formulating, selecting, and administering standards controlling design, construction, operations, and maintenance. The DOE Operations Office Manager was responsible for approving individual contractor practices and the governing site standards and requirements to be met. Due to the absence of comprehensive nuclear industry standards when most DOE sites were first established, Management and Operating contractors had to apply existing non-nuclear industry standards and, in many cases, formulate new technical standards to address unique applications

Equipment designs for the spent LWR fuel dry storage demonstration

International Nuclear Information System (INIS)

Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

1980-01-01

In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations

Testing round robin on cyclic crack growth of low and medium sulfur A533-B steels in LWR environments

International Nuclear Information System (INIS)

Kitagawa, H.; Komai, K.; Nakajima, H.; Higuchi, M.

1987-01-01

After the facts of environmentally assisted crack growth of low alloy steel was first observed when cyclically loaded in high temperature water. The subject has been extensively studied in connection with the evaluation of the integrity of LWR pressure boundary materials. In 1977, International Cooperative Group on Cyclic Crack Growth Rate Testing Evaluation (the ICCGR group) was organized for more systematic and effective solution of the problem. Successful results have been reported on the programs of the ICCGR activity, particularly in the promotion of a couple of programs of testing round robin and the associated research. JAERI also organized a domestic group of 15 organizations as the Corrosion Fatigue Subcommittee(JCF) of the LWR Safety Research Committee to carry out the similar test program. The group has been evaluating the behavior of steels representing the range of quality for the existing Japanese LWR plants. This paper describes the present status of the Japanese domestic testing round robin and related research especially focused on the test methodology

Study on reprocessing plant during transition period from LWR to FBR

International Nuclear Information System (INIS)

Shimada, Takashi; Matsui, Minefumi; Nishimura, Masashi; Ishida, Yasuhiro; Mori, Yukihide; Kuroda, Kazuhiko

2011-01-01

We have proposed a concept of a reprocessing plant suitable for the transition period from the light water reactors (LWRs) to the fast breeder reactors (FBRs) by making comparison of two plant concepts: (1) Independent Plant which processes LWR fuel and FBR fuel in separately constructed lines and (2) Modularized Plant which processes LWR fuel and FBR fuel in a same line. We made construction plans based on the reference power generation plan, and evaluated the Pu supply capability using the power generation plan as an indicator of plant operation flexibility. In general, a margin of processing capacity increases the Pu supply capability. The margin of the Modularized Plant necessary to obtain equivalent Pu supply capability is smaller than that of the Independent Plant. Also the margin of the Independent Plant results in decrease in the plant utilization factor. But the margin of the Modularized Plant results in little decrease in the plant utilization factor, because the Modularized Plant can address the types of reprocessing fuel to adjust to Pu demand and processing capacity. Therefore, the Modularized Plant has a greater potential for the reprocessing plants during transition period. (author)

The Width of High Burnup Structure in LWR UO2 Fuel

International Nuclear Information System (INIS)

Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong

2007-01-01

The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

Projection of US LWR spent fuel storage requirements

International Nuclear Information System (INIS)

Fletcher, J.F.; Cole, B.M.; Purcell, W.L.; Rau, R.G.

1982-11-01

The spent fuel storage requirements projection is based on data supplied for each operating or planned nuclear power power plant by the operting utilities. The data supplied by the utilities encompassed details of plant operating history, past records of fuel discharges, current inventories in reactor spent fuel storage pools, and projections of future discharge patterns. Data on storage capacity of storage pools and on characterization of the discharged fuel are also included. The data supplied by the utilities, plus additional data from other appropriate sources, are maintained on a computerized data base by Pacific Northwest Laboratory. The spent fuel requirements projection was based on utility data updated and verified as of December 31, 1981

50 CFR 23.20 - What CITES documents are required for international trade?

Science.gov (United States)

2010-10-01

... international trade? 23.20 Section 23.20 Wildlife and Fisheries UNITED STATES FISH AND WILDLIFE SERVICE..., EXPORTATION, AND IMPORTATION OF WILDLIFE AND PLANTS (CONTINUED) CONVENTION ON INTERNATIONAL TRADE IN... CITES documents are required for international trade? (a) Purpose. Articles III, IV, and V of the Treaty...

Determining if a change to a proposal requires additional NEPA documentation: the Smithsonian Solution; TOPICAL

International Nuclear Information System (INIS)

ECCLESTON, C.H.

1999-01-01

Proposed actions tend to evolve over time. Once National Environmental Policy Act (NEPA) documentation is completed, agencies are at risk that subsequent changes may not be adequately covered or that existing NEPA documentation maybe completely invalidated. Neither NEPA nor its subsequent regulations provide sufficient direction for determining the degree to which a proposed action may change before preparation of new or supplemental documentation is necessary. Yet, decisionmakers are routinely involved in determining if a change to a proposed action departs, to such an extent, from the description presented in the NEPA document that additional documentation is necessary. Experience demonstrates that no two decisionmakers will completely agree, one decisionmaker might believe that a particular change would not require additional documentation, while the other concludes the exact opposite. Lacking definitive direction, decisionmakers and critics alike may point to a universe of potential considerations as the basis for defending their claim that a change in an action does or does not require new or additional NEPA documentation. Assertions are often based on equivocal opinions that can be neither proved nor disproved. Moreover, decisionmakers are frequently placed in an arduous dilemma of justifying a decision, for which there is no generally accepted methodology on which to base the decision. Lack of definitive direction can prolong the decisionmaking process, resulting in project delays. This can also lead to inappropriate levels of NEPA documentation, inconsistencies in decisionmaking, and increased risk of a legal challenge because of insufficient documentation. Clearly, a more systematic and less subjective approach is needed, A tool for streamlining the NEPA process, by reducing this degree of subjectivity, is presented in this paper

Results of the LIRES Round Robin test on high temperature reference electrodes for LWR applications

Energy Technology Data Exchange (ETDEWEB)

Bosch, R.W. [SCK.CEN, Nuclear Research Centre Belgium, Boeretang 200, B-2400 Mol (Belgium); Nagy, G. [Magyar Tudomanyos Akademia KFKI Atomenergia Kutatointezet, AEKI, Konkoly Thege ut 29-33, 1121 Budapest (Hungary); Feron, D. [CEA Saclay, 91191 Gif-Sur-Yvette Cedex (France); Navas, M. [CIEMAT, Edificio 30, Dpto. Fision Nuclear, Avda. Complutense 22, 28040 Madrid, (Spain); Bogaerts, W. [KU Leuven, Kasteelpark Arenberg 31, B-3001 Leuven (Belgium); Karnik, D. [Nuclear Research Institute, NRI, Rez (Czech Republic); Dorsch, T. [Framatone ANP, Inc., Charlotte, North Carolina (United States); Molander, A. [Studsvik AB SE-611 82 Nykoeping (Sweden); Maekelae, K. [Materials and Structural Integrity, VTT Technical Research Centre of Finland, Kemistintie 3, P.O. Box 1704, FIN-02044 VTT (Finland)

2004-07-01

A European sponsored research project has been started on 1 October 2000 to develop high temperature reference electrodes that can be used for in-core electrochemical measurements in Light Water Reactors (LWR's). This LIRES-project (Development of Light Water Reactor Reference Electrodes) consists of 9 partners (SCK-CEN, AEKI, CEA, CIEMAT, KU Leuven, NRI Rez, Framatone ANP, Studsvik Nuclear and VTT) and will last for four years. The main objective of this LIRES project is to develop a reference electrode, which is robust enough to be used inside a LWR. Emphasize is put on the radiation hardness of both the mechanical design of the electrode as the proper functioning of the electrode. A four steps development trajectory is foreseen: (1) To set a testing standard for a Round Robin, (2) To develop different reference electrodes, (3) To perform a Round Robin test of these reference electrodes followed by selection of the best reference electrode(s), (4) To perform irradiation tests under appropriate LWR conditions in a Material Test Reactor (MTR). Four different high temperature reference electrodes have been developed and are being tested in a Round Robin test. These electrodes are: A Ceramic Membrane Electrode (CME), a Rhodium electrode, an external Ag/AgCl electrode and a Palladium electrode. The presentation will focus on the results obtained with the Round Robin test. (authors)

The necessity of improvement for the current LWR fuel assembly homogenization method

International Nuclear Information System (INIS)

Tang Chuntao; Huang Hao; Zhang Shaohong

2007-01-01

When the modern LWR core analysis method is used to do core nuclear design and in-core fuel management calculation, how to accurately obtain the fuel assembly homogenized parameters is a crucial issue. In this paper, taking the NEA C5G7-MOX benchmark problem as a severe test problem, which involves low-enriched uranium assemblies interspersed with MOX assemblies, we have re-examined the applicability of the two major assumptions of the modern equivalence theory for fuel assembly homoge- nization, i.e. the isolated assembly spatial spectrum assumption and the condensed two- group representation assumption. Numerical results have demonstrated that for LWR cores with strong spectrum interaction, both of these two assumptions are no longer applicable and the improvement for the homogenization method is necessary, the current two-group representation should be improved by the multigroup representation and the current reflective assembly boundary condition should be improved by the 'real' assembly boundary condition. This is a research project supported by National Natural Science Foundation of China (10605016). (authors)

MODRIB - a zero dimensional code for criticality and burn-up of LWR's

International Nuclear Information System (INIS)

Gaafar, M.A.; El-Cherif, A.I.

1980-01-01

The computer program MODRIB is a zero-dimensional code for calculating criticality and burn-up of light water reactors (LWR's). It is a version of an Italian code RIBOT-2 with an updated cross-section data library. The nuclear constants of MODRIB-code are calculated with a two group scheme (fast and thermal), where the fast group is an average of three fast groups. The code requires as input data essential extensive reactor parameters such as fuel rod radius, clad thickness, fuel enrichment, lattice pitch, water density and temperature etc. A summary of the physical model description and the input-output procedures are given in this report. Selected results of two sample problems are also given for the purpose of checking the validity and reliability of the code. The first is BWR and the second is PWR. The calculation time for a criticality problem with burn-up is about 8 seconds for the first time step and about 3 seconds for each subsequent time step on the ICL-1906 computer facility. The requirements on the memory size is less than 32 K-word. (author)

A fracture mechanics approach for estimating fatigue crack initiation in carbon and low-alloy steels in LWR coolant environments

International Nuclear Information System (INIS)

Park, H. B.; Chopra, O. K.

2000-01-01

A fracture mechanics approach for elastic-plastic materials has been used to evaluate the effects of light water reactor (LWR) coolant environments on the fatigue lives of carbon and low-alloy steels. The fatigue life of such steel, defined as the number of cycles required to form an engineering-size crack, i.e., 3-mm deep, is considered to be composed of the growth of (a) microstructurally small cracks and (b) mechanically small cracks. The growth of the latter was characterized in terms of ΔJ and crack growth rate (da/dN) data in air and LWR environments; in water, the growth rates from long crack tests had to be decreased to match the rates from fatigue S-N data. The growth of microstructurally small cracks was expressed by a modified Hobson relationship in air and by a slip dissolution/oxidation model in water. The crack length for transition from a microstructurally small crack to a mechanically small crack was based on studies on small crack growth. The estimated fatigue S-N curves show good agreement with the experimental data for these steels in air and water environments. At low strain amplitudes, the predicted lives in water can be significantly lower than the experimental values

Baseline descriptions for LWR spent fuel storage, handling, and transportation

Energy Technology Data Exchange (ETDEWEB)

Moyer, J.W.; Sonnier, C.S.

1978-04-01

Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables.

Baseline descriptions for LWR spent fuel storage, handling, and transportation

International Nuclear Information System (INIS)

Moyer, J.W.; Sonnier, C.S.

1978-04-01

Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables

Electronic document management meets environmental restoration recordkeeping requirements: A case study

International Nuclear Information System (INIS)

Burnham, S.L.

1995-01-01

Efforts at migrating records management at five Department of Energy sites operated under management by Lockheed Martin Energy Systems, Inc. for Environmental Restoration (ER) business activities are described. The corporate environment, project definition, records keeping requirements are described first. Then an evaluation of electronic document management technologies and of internal and commercially available systems are provided. Finally adopted incremental implementation strategy and lessons learned are discussed

Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

International Nuclear Information System (INIS)

Stout, R.B.; Merckx, K.R.; Holm, J.S.

1981-01-01

This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels

Performance of artificially defected LWR fuel rods in an unlimited air dry storage atmosphere

International Nuclear Information System (INIS)

Einziger, R.E.; Knecht, R.L.; Cantley, D.A.; Cook, J.A.

1983-09-01

Thus far the tests are inconclusive as to whether breached LWR fuel can be stored at 230 0 C for long periods of time in air without fuel oxidation and dispersion. There is every indication, as expected, that there is no oxidation problem in an inert atmosphere. Only one of four defects exposed to unlimited air gave any indication of fuel oxidation. It has been suggested that this might be an incubation effect and continued operation would result in oxidation occurring at all four defects. As yet the destructive examination of the BWR rod has not been completed, so it is not possible to determine if cladding splitting was due to an anomoly in this test rod or something that can be expected in LWR rods in general. Thus far there is no indication of respirable particle dispersal even if fuel oxidation does occur

Problems associated with domestic LWR technology development

International Nuclear Information System (INIS)

Watamori, Tikara

1975-01-01

To cope with the future energy problem in Japan, the enhancement of her own technology is continuing in the nuclear power field. Developments in the past, current state, and problems for the future are described regarding LWR power plants. The technology introduced from overseas countries cannot be used as it is. The domestic technology thus consists of the conversion of nuclear power technology so as to meet Japan's own condition and the domestic manufacture of machinery. In the former category, there are the aspects of aseismatic design, waste disposal, software, etc. In the latter, there are the productions of reactor vessels, steam generators, large valves, piping, etc. As the problems for the future, there are reliability and safety and the associated standardization. (Mori, K.)

Issues in risk analysis of passive LWR designs

International Nuclear Information System (INIS)

Youngblood, R.W.; Pratt, W.T.; Amico, P.J.; Gallagher, D.

1992-01-01

This paper discusses issues which bear on the question of how safety is to be demonstrated for ''simplified passive'' light water reactor (LWR) designs. First, a very simplified comparison is made between certain systems in today's plants. comparable systems in evolutionary designs, and comparable systems in the simplified passives. in order to introduce the issues. This discussion is not intended to describe the designs comprehensively, but is offered only to show why certain issues seem to be important in these particular designs. Next, an important class of accident sequences is described; finally, based on this discussion, some priorities in risk analysis are presented and discussed

Guidance for preparing user requirements documents for small and medium reactors and their application

International Nuclear Information System (INIS)

2000-08-01

During the past decade, several countries with highly developed nuclear power programs established user required documents (URDs) to guide the development and implementation of advanced light water reactors. These efforts built upon the extensive experience with operating reactors and included new insights from ongoing research and development to enhance the economic performance and safety of future nuclear power plants. Subsequently, a number of developing countries with plans for introducing nuclear energy into their national programs expressed strong interest in establishing analogous requirements. The IAEA has therefore taken the initiative to assist in the elaboration of such requirements. Building upon relevant documents this report recommends a URD structure and content outline to support developing countries in preparing their URDs for various applications of small and medium reactors (e.g. electricity generation and/or desalination). This report was prepared by representatives from both developing and developed Member States

Determination of Waste Isolation Pilot Plant (WIPP) management and institutional requirements documents for contact-handled (CH) critical systems

International Nuclear Information System (INIS)

1990-01-01

This document lists the critical requirements documents applicable to the receipt of contact-handled waste at the Waste Isolation Pilot Plant. It also describes the processes used to determine the applicability of each document. This analysis is based on the applicable documents that were in effect in the February 1988 time frame. 2 refs

Waste Management System Requirements Document

International Nuclear Information System (INIS)

1992-02-01

This DCP establishes an interim plan for the Office of Civilian Radioactive Waste Management (OCRWM) technical baseline until the results of the OCRWM Document Hierarchy Task Force can be implemented. This plan is needed to maintain continuity in the Program for ongoing work in the areas of Waste Acceptance, Transportation, Monitored Retrievable Storage (MRS) and Yucca Mountain Site Characterization

Overview of LWR severe accident research activities at the Karlsruhe Institute of Technology

International Nuclear Information System (INIS)

Miassoedov, Alexei; Albrecht, Giancarlo; Foit, Jerzy-Jan; Jordan, Thomas; Steinbrück, Martin; Stuckert, Juri; Tromm, Walter

2012-01-01

The research activities in the light water reactor (LWR) severe accidents domain at Karlsruhe Institute of Technology (KIT) are concentrated on the in- and ex-vessel core melt behavior. The overall objective is to investigate the core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity and to the containment, corium concrete interaction and corium coolability in the reactor cavity, and hydrogen behaviour in reactor systems. The results of the experiments contribute to a better understanding of the core melt sequences and thus improve safety of existing and, in the long-term, of future reactors by severe accident mitigation measures and by safety installations where required. This overview paper describes the experimental facilities used at KIT for severe accident research and gives an overview of the main directions and objectives of the R&D work. (author)

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 7. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Burt, D.L.

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 7) presents the standards and requirements for the following sections: Occupational Safety and Health, and Environmental Protection.

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 7. Revision 1

International Nuclear Information System (INIS)

Burt, D.L.

1994-04-01

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 7) presents the standards and requirements for the following sections: Occupational Safety and Health, and Environmental Protection

Emergency Response Capability Baseline Needs Assessment - Requirements Document

Energy Technology Data Exchange (ETDEWEB)

Sharry, J A

2016-10-04

This document was prepared by John A. Sharry, LLNL Fire Marshal and LLNL Division Leader for Fire Protection and reviewed by LLNL Emergency Management Department Head James Colson. The document follows and expands upon the format and contents of the DOE Model Fire Protection Baseline Capabilities Assessment document contained on the DOE Fire Protection Web Site, but only addresses emergency response.

Long-term embrittlement of cast duplex stainless steels in LWR systems

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.

1990-08-01

This progress report summarizes work performed by Argonne National Laboratory on long-term embrittlement of cast duplex stainless steels in LWR systems during the six months from April to September 1988. Characteristics of the primary mechanism of aging embrittlement (i.e., spinodal decomposition of ferrite) and synergistic effects of alloying and impurity elements that influence the kinetics of the primary mechanism are discussed. Several secondary metallurgical processes of embrittlement, strongly dependent on the C, N, Ni, Mo, and Si content of various heats, are identified. Information on kinetics and data on impact properties are analyzed and correlated with microstructural characteristics to provide a unified method of extrapolating accelerated-aging data to reactor operating conditions. Fracture toughness data are presented for several heats of cast stainless steel aged at temperatures between 320 and 450 degrees C for times up to 10,000 h. Mechanical property data are analyzed to develop the procedure and correlations or predicting the kinetics and extent of embrittlement of reactor components from known material parameters. The method and examples of estimating the impact strength and fracture toughness of cast components during reactor service are described. The lower-bound values of impact strength and fracture toughness for cast stainless steels at LWR operating temperatures are defined. 42 refs., 14 figs., 6 tabs

Near-term objectives of the works on the EUR document, comparison with the EPRI-URD

International Nuclear Information System (INIS)

Berbey, P.; Lopez, P. T. L.

2010-01-01

18 years after its foundation, the EUR organization is still quite active. During the last 3 years, the EUR products have mainly been evaluations of the Gen 3 LWR designs. The evaluations of the AP1000 and of the AES92 designs have been concluded in 2007 and a revised version of the evaluation of the EPR completed in 2009. Other LWR projects of potential interest for the EUR utilities, such as MHI's APWR are being reviewed before starting a full-scope assessment. Last, a revision C of the EUR volume 4 has been published in 2007. Coordinated actions with the other industry groups and the other stakeholders have been a centerpiece of the recent EUR strategy. In particular, the EUR and ENISS organizations have joined their efforts to coordinate their actions in nuclear safety, in particular vs. IAEA and WENRA. Also EUR and WNA/CORDEL have coordinated their efforts, in relation with the MDEP initiative of the safety regulators. Meanwhile, the EUR organization has kept enlarging: CEZ and MVM now are associated members and Gen-Energija from Slovenia has been invited to participate. Education and training has been dealt with actively. The EUR organization strongly supported WNU in setting up their successful 2009 'forum on harmonization'. A more technical course about the EUR requirements is being organized under the aegis of ENEN. Finally, preparatory material for a revision D of the EUR volumes 1 and 2 has been gathered. Out of this material, the update of the comparison between the EUR document and the EPRI-URD is of specific importance. In 1995-1997 a first analysis of the differences between the two documents had been carried out by the EUR organization and a synthesis report focused on 35 'macro issues' had been produced. Ten years later, this work was felt deserving a thorough update. Thus a specific organization has been set out between EUR and EPRI to update the old synthesis report. The 'macro issues' have been re-analyzed by a specific EUR working group leaded

Development and application of best-estimate LWR safety analysis codes

International Nuclear Information System (INIS)

Reocreux, M.

1997-01-01

This paper is a review of the status and the future orientations of the development and application of best estimate LWR safety analysis codes. The present status of these codes exhibits a large success and almost a complete fulfillment of the objectives which were assigned in the 70s. The applications of Best Estimate codes are numerous and cover a large variety of safety questions. However these applications raised a number of problems. The first ones concern the need to have a better control of the quality of the results. This means requirements on code assessment and on uncertainties evaluation. The second ones concern needs for code development and specifically regarding physical models, numerics, coupling with other codes and programming. The analysis of the orientations for code developments and applications in the next years, shows that some developments should be made without delay in order to solve today questions whereas some others are more long term and should be tested for example in some pilot programmes before being eventually applied in main code development. Each of these development programmes are analyzed in the paper by detailing their main content and their possible interest. (author)

Standardization of dosimetry and damage analysis work for U.S. LWR, FBR, and MFR development program

International Nuclear Information System (INIS)

McElroy, W.N.; Doran, D.G.; Gold, R.; Morgan, W.C.; Grundl, J.A.; McGarry, E.D.; Kam, F.B.K.; Swank, J.H.; Odette, G.R.

1978-01-01

The accuracy requirements for various measured/calculated exposure and correlation parameters associated with current dosimetry and damage analysis procedures and practices depend on the accuracy needs of reactor development efforts in testing, design, safety, operations, and surveillance programs. Present state-of-the-art accuracies are estimated to be in the range of +-2 to 30 percent (1 sigma), depending on the particular parameter. There now appears to be international agreement, at least for the long term, that most reactor fuels and materials programs will not be able to accept an uncertainty greater than about +5 percent (1 sigma). The current status of dosimetry and damage analysis standardization work within the U.S. for LWR, FBR and MFR is reviewed in this paper

Total sensitivity and uncertainty analysis for LWR pin-cells with improved UNICORN code

International Nuclear Information System (INIS)

Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei

2017-01-01

Highlights: • A new model is established for the total sensitivity and uncertainty analysis. • The NR approximation applied in S&U analysis can be avoided by the new model. • Sensitivity and uncertainty analysis is performed to PWR pin-cells by the new model. • The effects of the NR approximation for the PWR pin-cells are quantified. - Abstract: In this paper, improvements to the multigroup cross-section perturbation model have been proposed and applied in the self-developed UNICORN code, which is capable of performing the total sensitivity and total uncertainty analysis for the neutron-physics calculations by applying the direct numerical perturbation method and the statistical sampling method respectively. The narrow resonance (NR) approximation was applied in the multigroup cross-section perturbation model, implemented in UNICORN. As improvements to the NR approximation to refine the multigroup cross-section perturbation model, an ultrafine-group cross-section perturbation model has been established, in which the actual perturbations are applied to the ultrafine-group cross-section library and the reconstructions of the resonance cross sections are performed by solving the neutron slowing-down equation. The total sensitivity and total uncertainty analysis were then applied to the LWR pin-cells, using both the multigroup and the ultrafine-group cross-section perturbation models. The numerical results show that the NR approximation overestimates the relative sensitivity coefficients and the corresponding uncertainty results for the LWR pin-cells, and the effects of the NR approximation are significant for σ_(_n_,_γ_) and σ_(_n_,_e_l_a_s_) of "2"3"8U. Therefore, the effects of the NR approximation applied in the total sensitivity and total uncertainty analysis for the neutron-physics calculations of LWR should be taken into account.

Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

International Nuclear Information System (INIS)

Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

2001-01-01

The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

New development in nondestructive measurement and verification of irradiated LWR fuels

International Nuclear Information System (INIS)

Lee, D.M.; Phillips, J.R.; Halbig, J.K.; Hsue, S.T.; Lindquist, L.O.; Ortega, E.M.; Caine, J.C.; Swansen, J.; Kaieda, K.; Dermendjiev, E.

1979-01-01

Nondestructive techniques for characterizing irradiated LWR fuel assemblies are discussed. This includes detection systems that measure the axial activity profile, neutron yield and gamma yield. A multi-element profile monitor has been developed that offers a significant improvement in speed and complexity over existing mechanical scanning systems. New portable detectors and electronics, applicable to safeguard inspection, are presented and results of gamma-ray and neutron measurements at commercial reactor facilities are given

Validation of Monte Carlo predictions of LWR-PROTEUS safety parameters using an improved whole-reactor model

Energy Technology Data Exchange (ETDEWEB)

Plaschy, M. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, CH-5232 Villigen, PSI (Switzerland)], E-mail: michael.plaschy@eos.ch; Murphy, M.; Jatuff, F.; Perret, G.; Seiler, R. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, CH-5232 Villigen, PSI (Switzerland); Chawla, R. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, CH-5232 Villigen, PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne, EPFL (Switzerland)

2009-10-15

The recent experimental programme conducted in the PROTEUS research reactor at the Paul Scherrer Institute (PSI) has concerned detailed investigations of advanced light water reactor (LWR) fuels. More than fifteen different configurations of the multi-zone critical facility have been studied, each of them requiring accurate estimation of operational safety parameters, in particular the critical driver loadings, shutdown rod worths and the effective delayed neutron fraction {beta}{sub eff}. The current paper presents a full-scale 3D Monte Carlo model for the facility, set up using the MCNPX code, which has been employed for calculation of the operational characteristics for seven different LWR-PROTEUS configurations. Thereby, a variety of nuclear data libraries (viz. ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JEFF3.1, JENDL3.2, and JENDL3.3) have been used, and predictions of k{sub eff} and shutdown rod worths compared with experimental values. Even though certain library-specific trends have been observed, the k{sub eff} predictions are generally very satisfactory, viz. with discrepancies of <0.5% between calculation (C) and experiment (E). The results also confirm the consistent determination of reactivity variations, the C/E values for the shutdown (safety) rod worths being always within 5% of unity. In addition, the MCNP modelling of the multi-zone reactor has yielded interesting results for the delayed neutron fraction ({beta}{sub eff}) in the different configurations, a breakdown being made possible in each case in terms of delayed neutron group, fissioning nuclide, and reactor region.

Standard casks for the transport of LWR spent fuel

International Nuclear Information System (INIS)

Blum, P.

1985-01-01

During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufacturer under TRANSNUCLEAIRE supervision in different countries and are presently used for European and intercontinental transports. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardisation facilitating fabrication, operation and spare part supply [fr

Hamor-2: a computer code for LWR inventory calculation

International Nuclear Information System (INIS)

Guimaraes, L.N.F.; Marzo, M.A.S.

1985-01-01

A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt

Conceptual design of a spent LWR fuel recycle complex

International Nuclear Information System (INIS)

Kirk, B.H.

1980-01-01

Purpose was to design a licensable facility, to make cost-benefit analyses of alternatives, and to aid in developing licensing criteria. The Savannah River Plant was taken to be the site for the recycle complex. The spent LWR fuel will be processed through the plant at the rate of 3000 metric tons of heavy metal per year. The following aspects of the complex are discussed: operation, maintenance, co-conversion (Coprecal), waste disposal, off-gas treatment, ventilation, safeguards, accounting, equipment and fuel fabrication. Differences between the co-processing case and the separated streams case are discussed. 44 figures

Integrity of neutron-absorbing components of LWR fuel systems

International Nuclear Information System (INIS)

Bailey, W.J.; Berting, F.M.

1991-03-01

A study of the integrity and behavior of neutron-absorbing components of light-water (LWR) fuel systems was performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE). The components studies include control blades (cruciforms) for boiling-water reactors (BWRs) and rod cluster control assemblies for pressurized-water reactors (PWRs). The results of this study can be useful for understanding the degradation of neutron-absorbing components and for waste management planning and repository design. The report includes examples of the types of degradation, damage, or failures that have been encountered. Conclusions and recommendations are listed. 84 refs

Measurement and characterization of fission products released from LWR fuel

International Nuclear Information System (INIS)

Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

1984-01-01

Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from proportional 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag. (orig./HP)

Spent fuel storage requirements. An update of DOE/RL-83-1

International Nuclear Information System (INIS)

1984-05-01

Spent fuel storage capacities at some commercial light water reactors (LWRs) are inadequate to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information voluntarily supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) Federal Interim Storage (FIS) Program and the spent fuel research, development, and demonstration (RD and D) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. Since the planning needs of the CSFM program focus on the near-term management of spent fuel inventories from commercial nuclear power reactors, the estimates in this report cover the ten-year period from the present through 1983. The report also assesses the possible impacts of using various concepts to reduce the requirements for additional storage capacity

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

Energy Technology Data Exchange (ETDEWEB)

Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-01

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

International Nuclear Information System (INIS)

Honma, George

2015-01-01

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

Transmutation of LWR waste actinides in thermal reactors

International Nuclear Information System (INIS)

Gorrell, T.C.

1979-01-01

Recycle of actinides to a reactor for transmutation to fission products is being considered as a possible means of waste disposal. Actinide transmutation calculations were made for two irradiation options in a thermal (LWR) reactor. The cases considered were: all actinides recycled in regular uranium fuel assemblies, and transuranic actinides recycled in separate mixed oxide (MOX) assemblies. When all actinides were recycled in a uranium lattice, a reduction of 62% in the transuranic inventory was achieved after 10 recycles, compared to the inventory accumulated without recycle. When the transuranics from 2 regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after 5 recycles

Validating the BISON fuel performance code to integral LWR experiments

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

2016-05-15

Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 7

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

This Requirements Identification Document (RID) describes an Occupational Health and Safety Program as defined through the Relevant DOE Orders, regulations, industry codes/standards, industry guidance documents and, as appropriate, good industry practice. The definition of an Occupational Health and Safety Program as specified by this document is intended to address Defense Nuclear Facilities Safety Board Recommendations 90-2 and 91-1, which call for the strengthening of DOE complex activities through the identification and application of relevant standards which supplement or exceed requirements mandated by DOE Orders. This RID applies to the activities, personnel, structures, systems, components, and programs involved in maintaining the facility and executing the mission of the High-Level Waste Storage Tank Farms.

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 7

International Nuclear Information System (INIS)

1994-04-01

This Requirements Identification Document (RID) describes an Occupational Health and Safety Program as defined through the Relevant DOE Orders, regulations, industry codes/standards, industry guidance documents and, as appropriate, good industry practice. The definition of an Occupational Health and Safety Program as specified by this document is intended to address Defense Nuclear Facilities Safety Board Recommendations 90-2 and 91-1, which call for the strengthening of DOE complex activities through the identification and application of relevant standards which supplement or exceed requirements mandated by DOE Orders. This RID applies to the activities, personnel, structures, systems, components, and programs involved in maintaining the facility and executing the mission of the High-Level Waste Storage Tank Farms

Physicochemical characterization of solidification agents used and products formed with radioactive wastes at LWR nuclear power plants

International Nuclear Information System (INIS)

Kibbey, A.H.; Godbee, H.W.

1978-01-01

Solidification of evaporator concentrates, filter sludges, and spent ion exchange resins used in LWR streams is discussed. The introduction of solidification agents to immobilize these sludges and resins can increase the volume of these wastes by a factor of slightly over 1 to greater than 2, depending on the binder chosen. The agents and methods used or proposed for use in solidification of LWR power plant wastes are generally suitable for treating most of the other-than-high-level wastes generated throughout the entire fuel cycle. Among the solidification agents most commonly used or suggested for use are the inorganic cements and organic plastics, which are listed and compared. A summary of considerations important in choosing a solidification agent is presented tabularly

20 CFR 669.660 - What planning documents and information are required in the application for MSFW youth grants and...

Science.gov (United States)

2010-04-01

... 20 Employees' Benefits 3 2010-04-01 2010-04-01 false What planning documents and information are... PROGRAM UNDER TITLE I OF THE WORKFORCE INVESTMENT ACT The MSFW Youth Program § 669.660 What planning...? The required planning documents and other required information and the submission dates for filing are...

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Feasibility study on the development of advanced LWR fuel technology

International Nuclear Information System (INIS)

Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others.

1997-07-01

Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO 2 pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO 2 -Gd 2 O 3 burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs

Feasibility study on the development of advanced LWR fuel technology

Energy Technology Data Exchange (ETDEWEB)

Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

1997-07-01

Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

LMBFR and LWR in-core thermal-hydraulic codes: the state-of-the-art and research and development needs

International Nuclear Information System (INIS)

Khan, E.U.; Coomes, E.P.; Rowe, D.S.; Trent, D.S.

1981-04-01

A review of analytical design methods used for predicting reactor core flow and temperature distributions is presented with emphasis on LMFBR's. The paper also briefly describes and contrasts the methods used for LWR's. These methods are global analysis, subchannel analysis, distributed parameter, and hybrid analysis. The evolution of the local and subchannel analysis methods is presented. Data used for code validation are also presented. Current research and development needs are identified and discussed. Areas identified for future research and development include methods and expermental data for analysis of distorted bundles and natural convection. Methods that have been developed for predicting the safety performance of LMFBR's and LWR's are not within the scope of this paper

Fast reactor core design studies to cope with TRU fuel composition changes in the LWR-to-FBR transition period

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

2009-01-01

As part of the Fast Reactor Cycle Technology Development Project (FaCT Project), sodium-cooled fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. A 750 MWe mixed-oxide fuel core is firstly defined as a FaCT medium-size reference core and its neutronics characteristics are determined. The core is a high internal conversion type and has an average burnup of 150 GWD/T. The reference TRU fuel composition is assumed to come from the FBR equilibrium state. Compared to the LWR-to-FBR transition period, the TRU fuels in the FBR equilibrium period are multi-recycled through fast reactors and have a different composition. An available TRU fuel composition is determined by fast reactor spent fuel multi-recycling scenarios. Then the FaCT core corresponding to the TRU fuel with different compositions is set according to the TRU fuel composition changes in LWR-to-FBR transition period, and the key core neutronics characteristics are assessed. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters. (author)

Ice Thermal Storage Systems for LWR Supplemental Cooling and Peak Power Shifting

Energy Technology Data Exchange (ETDEWEB)

Haihua Zhao; Hongbin Zhang; Phil Sharpe; Blaise Hamanaka; Wei Yan; WoonSeong Jeong

2010-06-01

Availability of enough cooling water has been one of the major issues for the nuclear power plant site selection. Cooling water issues have frequently disrupted the normal operation at some nuclear power plants during heat waves and long draught. The issues become more severe due to the new round of nuclear power expansion and global warming. During hot summer days, cooling water leaving a power plant may become too hot to threaten aquatic life so that environmental regulations may force the plant to reduce power output or even temporarily to be shutdown. For new nuclear power plants to be built at areas without enough cooling water, dry cooling can be used to remove waste heat directly into the atmosphere. However, dry cooling will result in much lower thermal efficiency when the weather is hot. One potential solution for the above mentioned issues is to use ice thermal storage systems (ITS) that reduce cooling water requirements and boost the plant’s thermal efficiency in hot hours. ITS uses cheap off-peak electricity to make ice and uses those ice for supplemental cooling during peak demand time. ITS is suitable for supplemental cooling storage due to its very high energy storage density. ITS also provides a way to shift large amount of electricity from off peak time to peak time. Some gas turbine plants already use ITS to increase thermal efficiency during peak hours in summer. ITSs have also been widely used for building cooling to save energy cost. Among three cooling methods for LWR applications: once-through, wet cooling tower, and dry cooling tower, once-through cooling plants near a large water body like an ocean or a large lake and wet cooling plants can maintain the designed turbine backpressure (or condensation temperature) during 99% of the time; therefore, adding ITS to those plants will not generate large benefits. For once-through cooling plants near a limited water body like a river or a small lake, adding ITS can bring significant economic

Comment: collection of assay data on isotopic composition in LWR spent fuel

Energy Technology Data Exchange (ETDEWEB)

Naito, Yoshitaka; Kurosawa, Masayoshi; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

Many assay data of LWR spent fuels have been collected from reactors in the world and some of them are already stored in the database SFCOMPO which was constructed on a personal computer IBM PC/AT. On the other hand, Group constant libraries for burnup calculation code ORIGEN-II were generated from the nuclear data file JENDL3.2. These libraries were evaluated by using the assay data in SFCOMPO. (author)

Grid requirements applicable to future NPPs in the new European Electricity Framework

International Nuclear Information System (INIS)

Beato Castro, D.; Padill, C. M.

2000-01-01

With a view to keeping nuclear energy as an option for future power generation in a competitive market and taking advantage of the current operating experience, a group of European electric utilities have come together to define common requirements for the design and supply of future Light Water Reactor (LWR) plants connected to the electrical system. These requirements, defined with the aim of standardizing and adapting design to the conditions of the new electricity framework, are being included in the European Utility Requirements (EUR) document. Although there are different types of power plants operating appropriately in large electrical systems, the idea is to find the minimum requirements that will allow growth of this type of energy in the European electricity industry without reducing quality, safety and reliability of interconnected electrical systems. It is therefore necessary to take into account the features of the existing power systems and the operating characteristics and design of nuclear power plants so as to harmonize their respective technical peculiarities in the framework of the deregulated electricity sector. The definition of these grid requirements is based primarily on the operating conditions of the Union pour la Coordination de la Production et le Transport de L'Electricite (UCPTE) grid and takes into account the current Grid Code of the main European countries, for the forthcoming Issue C. This paper sets outs the most relevant aspects of the grid requirements, included in Chapter 2.3 of the EUR document Grid Requirements, Issue B, for the connection of future nuclear power plants in the European electricity system, and others that are being considered in the preparation of the new issue of the document that will take into account the deregulated electricity market situation and deal with the following aspects: General characteristics. Operation of a plant under normal grid conditions. Operation of a plant under disturbed grid

Stylized whole-core benchmark of the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept

International Nuclear Information System (INIS)

Hon, Ryan; Kooreman, Gabriel; Rahnema, Farzad; Petrovic, Bojan

2017-01-01

Highlights: • A stylized benchmark specification of the I2S-LWR core. • A library of cross sections were generated in both 8 and 47 groups. • Monte Carlo solutions generated for the 8 group library using MCNP5. • Cross sections and pin fission densities provided in journal’s repository. - Abstract: The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) is a pressurized water reactor (PWR) concept under development by a multi-institutional team led by Georgia Tech. The core is similar in size to small 2-loop PWRs while having the power level of current large reactors (∼1000 MWe) but using uranium silicide fuel and advanced stainless steel cladding. A stylized benchmark specification of the I 2 S-LWR core has been developed in order to test whole-core neutronics codes and methods. For simplification the core was split into 57 distinct material regions for cross section generation. Cross sections were generated using the lattice physics code HELIOS version 1.10 in both 8 and 47 groups. Monte Carlo solutions, including eigenvalue and pin fission densities, were generated for the 8 group library using MCNP5. Due to space limitations in this paper, the full cross section library and normalized pin fission density results are provided in the journal’s electronic repository.

Technical program to study the benefits of nonlinear analysis methods in LWR component designs. Technical report TR-3723-1

International Nuclear Information System (INIS)

Raju, P.P.

1980-05-01

This report summarizes the results of the study program to assess the benefits of nonlinear analysis methods in Light Water Reactor (LWR) component designs. The current study reveals that despite its increased cost and other complexities, nonlinear analysis is a practical and valuable tool for the design of LWR components, especially under ASME Level D service conditions (faulted conditions) and it will greatly assist in the evaluation of ductile fracture potential of pressure boundary components. Since the nonlinear behavior is generally a local phenomenon, the design of complex components can be accomplished through substructuring isolated localized regions and evaluating them in detail using nonlinear analysis methods

42 CFR 495.338 - Health information technology implementation advance planning document requirements (HIT IAPD).

Science.gov (United States)

2010-10-01

... 42 Public Health 5 2010-10-01 2010-10-01 false Health information technology implementation... CERTIFICATION STANDARDS FOR THE ELECTRONIC HEALTH RECORD TECHNOLOGY INCENTIVE PROGRAM Requirements Specific to the Medicaid Program § 495.338 Health information technology implementation advance planning document...

Pie technique of LWR fuel cladding fracture toughness test

International Nuclear Information System (INIS)

Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu

2006-01-01

Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)

Alternatives for managing post LWR reactor nuclear wastes

International Nuclear Information System (INIS)

Platt, A.M.

1976-01-01

The two extremes in the LWR fuel cycle are discarding the spent fuel and recycling the U and Pu to the maximum extent possible. The waste volumes from the two alternatives are compared. A preliminary evaluation is made of the technology available for handling wastes from each step of the fuel cycle. The wastes considered are fuel materials, high--level wastes, other liquids, combustible and non-combustible solids, and non--high--level wastes. Evaluation of processing gaseous wastes indicates that technology is available for capture of Kr and I 2 , but further development is needed for T 2 . Technology for interim storage and geological isolation is considered adequate. An outline is given of the steps in the selection of a final storage site

Application of an enhanced cross-section interpolation model for highly poisoned LWR core calculations

International Nuclear Information System (INIS)

Palau, J.M.; Cathalau, S.; Hudelot, J.P.; Barran, F.; Bellanger, V.; Magnaud, C.; Moreau, F.

2011-01-01

Burnable poisons are extensively used by Light Water Reactor designers in order to preserve the fuel reactivity potential and increase the cycle length (without increasing the uranium enrichment). In the industrial two-steps (assembly 2D transport-core 3D diffusion) calculation schemes these heterogeneities yield to strong flux and cross-sections perturbations that have to be taken into account in the final 3D burn-up calculations. This paper presents the application of an enhanced cross-section interpolation model (implemented in the French CRONOS2 code) to LWR (highly poisoned) depleted core calculations. The principle is to use the absorbers (or actinide) concentrations as the new interpolation parameters instead of the standard local burnup/fluence parameters. It is shown by comparing the standard (burnup/fluence) and new (concentration) interpolation models and using the lattice transport code APOLLO2 as a numerical reference that reactivity and local reaction rate prediction of a 2x2 LWR assembly configuration (slab geometry) is significantly improved with the concentration interpolation model. Gains on reactivity and local power predictions (resp. more than 1000 pcm and 20 % discrepancy reduction compared to the reference APOLLO2 scheme) are obtained by using this model. In particular, when epithermal absorbers are inserted close to thermal poison the 'shadowing' ('screening') spectral effects occurring during control operations are much more correctly modeled by concentration parameters. Through this outstanding example it is highlighted that attention has to be paid to the choice of cross-section interpolation parameters (burnup 'indicator') in core calculations with few energy groups and variable geometries all along the irradiation cycle. Actually, this new model could be advantageously applied to steady-state and transient LWR heterogeneous core computational analysis dealing with strong spectral-history variations under

Parameters' influence estimation on Puf supply and demand in transitional period from LWR to FBR in Japan

International Nuclear Information System (INIS)

Kobayashi, Hiroaki; Ohta, Hirokazu; Inoue, Tadashi

2009-01-01

Plutonium fissile (Puf) amounts to balance supply and demand during transition period were evaluated with different parameters. Estimated total Puf demand in transitional period was sensitive to deployment speed of FBR. Because FBRs will be deployed as replacements of old LWRs for keeping total capacity, deployment history of existing LWRs should be taken into consideration. According to the estimation, LWR fuel burnup and utilized capacity are not big issue. Because certain amount of LWR spent fuel will remain in early phase of transitional period, there is enough time for preparing Puf supply. On the other hand, FBR fuel cycle time (SF cooling time + fuel fabrication time) have large impact on Puf supply. Fuel cycle technologies including transportation for applying to short cooling spent fuels should be developed. (author)

Investigation of nuclide importance to functional requirements related to transport and long-term storage of LWR spent fuel

International Nuclear Information System (INIS)

Broadhead, B.L.; DeHart, M.D.; Ryman, J.C.; Tang, J.S.; Parks, C.V.

1995-06-01

This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to 100,000 years. Ranking plots for each of these analysis areas are given in an Appendix for completeness, as well as summary tables in the main body of the report. Summary rankings are presented in terms of high (greater than 10% contribution to the total), medium (between 1% and 10% contribution), and low (less than 1% contribution) for both short- and long-term cooling. When compared with the expected measurement accuracies, these rankings show that most of the important isotopes can be characterized sufficiently for the purpose of radionuclide generation/depletion code validation in each of the analysis areas. Because the main focus of this work is on the relative importances of isotopes associated with L at sign spent fuel, some conclusions may not be applicable to similar areas such as high-level waste (HLW) and nonfuel-bearing components (NFBC)

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Preliminary concepts for detecting diversion of LWR spent fuel

International Nuclear Information System (INIS)

Sellers, T.A.

Sandia Laboratories, under the sponsorship of the Department of Energy, Office of Safeguards and Security, has been developing conceptual designs of advanced systems to rapidly detect diversion of LWR spent fuel. Three detection options have been identified and compared on the basis of timeliness of detection and cost. Option 1 is based upon inspectors visiting each facility on a periodic basis to obtain and review data acquired by surveillance instruments and to verify the inventory. Option 2 is based upon continuous inspector presence, aided by surveillance instruments. Option 3 is based upon the collection of data from surveillance instruments with periodic readout either at the facility or at a remote central monitoring and display module and occasional inspection. Surveillance instruments are included in each option to assure a sufficiently high probability of detection. An analysis technique with an example logic tree that was used to identify performance requirements is described. A conceptual design has been developed for Option 3 and the essential hardware elements are not being developed. These elements include radiation, crane and pool acoustic sensors, a Data Collection Module, a Local Collection Module, a Local Display Module and a Central Monitoring and Display Module. A demonstration, in operating facilities, of the overall system concept is planned for the March to June 1979 time frame

Functional requirements document for NASA/MSFC Earth Science and Applications Division: Data and information system (ESAD-DIS). Interoperability, 1992

Science.gov (United States)

Stephens, J. Briscoe; Grider, Gary W.

1992-01-01

These Earth Science and Applications Division-Data and Information System (ESAD-DIS) interoperability requirements are designed to quantify the Earth Science and Application Division's hardware and software requirements in terms of communications between personal and visualization workstation, and mainframe computers. The electronic mail requirements and local area network (LAN) requirements are addressed. These interoperability requirements are top-level requirements framed around defining the existing ESAD-DIS interoperability and projecting known near-term requirements for both operational support and for management planning. Detailed requirements will be submitted on a case-by-case basis. This document is also intended as an overview of ESAD-DIs interoperability for new-comers and management not familiar with these activities. It is intended as background documentation to support requests for resources and support requirements.

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Measurement and characterization of fission products released from LWR fuel

International Nuclear Information System (INIS)

Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

1984-01-01

Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag

Spent fuel storage requirements. An update of DOE/RL-85-2

International Nuclear Information System (INIS)

1986-10-01

Utility projections of spent fuel storage capacities indicate that some commercial light water reactors (LWRs) have inadequate capacity to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. The estimates in this report cover the period from the present through the year 2000. Although the DOE objective is to begin accepting spent fuel for final disposal in 1998, types of fuel and the receipt rates to be shipped are not yet known. Hence, this report makes no assumption regarding such fuel shipments. The report also assesses the possible impacts of increased fuel exposure and spent fuel transshipment on the requirements for additional storage capacity

Instructions for submittal and control of FFTF design documents and design related documentation

International Nuclear Information System (INIS)

Grush, R.E.

1976-10-01

This document provides the system and requirements for management of FFTF technical data prepared by Westinghouse Hanford (HEDL), and design contractors, the construction contractor and lower tier equipment suppliers. Included in this document are provisions for the review, approval, release, change control, and accounting of FFTF design disclosure and base documentation. Also included are provisions for submittal of other design related documents for review and approval consistent with applicable requirements of RDT-Standard F 2-2, ''Quality Assurance Program Requirements.''

I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology with SCALE6.1

International Nuclear Information System (INIS)

Matijevic, M.; Pevec, D.; Jecmenica, R.

2016-01-01

The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept developed by Georgia Tech is a novel PWR reactor delivering electric power of 1000 MWe while implementing inherent safety features typical for Generation III+ small modular reactors. The main safety feature is based on integral primary circuit configuration, bringing together compact design of the reactor core with 121 fuel assembly (FA), control rod drive mechanism (CRDM), 8 primary heat exchangers (PHE), 4 passive decay heat removal systems (DHRS), 8 pumps, and other integral components. A high power density core based on silicide fuel is selected to achieve a high thermal power which is extracted with PHEs placed in the annual region between the barrel and the vessel. The complex and integrated design of I2S-LWR leads to activation of integral components, mainly made from stainless steel, so accurate and precise Monte Carlo (MC) simulations are needed to quantify potential dose rates to personnel during routine maintenance operation. This shielding problem is therefore very challenging one, posing a non-trivial neutron flux solution in a phase space. This paper presents the performance of the hybrid shielding methodologies CADIS/FW-CADIS implemented in the MAVRIC sequence of the SCALE6.1 code package. The main objective was to develop a detailed MC shielding model of the I2S-LWR reactor along with effective variance reduction (VR) parameters and to calculate neutron fluence rates inside PHEs. Such results are then utilized to find neutron activation rate distribution via 60Co generation inside of a stack of microchannel heat exchangers (MCHX), which will be periodically withdrawn for the maintenance. 59Co impurities are the main cause of (n,gamma) radiative gamma dose to personnel via neutron activation since 60Co has half-life of 5.27 years and is emitting high energy gamma rays (1.17 MeV and 1.33 MeV). The developed MC model was successfully used to find converged fluxes inside all 8 stacks of

On the Diffusion Coefficient of Two-step Method for LWR analysis

International Nuclear Information System (INIS)

Lee, Deokjung; Choi, Sooyoung; Smith, Kord S.

2015-01-01

The few-group constants including diffusion coefficients are generated from the assembly calculation results. Once the assembly calculation is done, the cross sections (XSs) are spatially homogenized, and a critical spectrum calculation is performed in order to take into account the neutron leakages of the lattice. The diffusion coefficient is also generated through the critical spectrum calculation. Three different methods of the critical spectrum calculation such as B1 method, P1 method, and fundamental mode (FM) calculation method are considered in this paper. The diffusion coefficients can also be affected by transport approximations for the transport XS calculation which is used in the assembly transport lattice calculation in order to account for the anisotropic scattering effects. The outflow transport approximation and the inflow transport approximation are investigated in this paper. The accuracy of the few group data especially the diffusion coefficients has been studied to optimize the combination of the transport correction methods and the critical spectrum calculation methods using the UNIST lattice physics code STREAM. The combination of the inflow transport approximation and the FM method is shown to provide the highest accuracy in the LWR core calculations. The methodologies to calculate the diffusion coefficients have been reviewed, and the performances of them have been investigated with a LWR core problem. The combination of the inflow transport approximation and the fundamental mode critical spectrum calculation shows the smallest errors in terms of assembly power distribution

Performance and reliability of LWR fuel

International Nuclear Information System (INIS)

Bairiot, H.; Deramaix, P.; Vandenberg, C.

1977-01-01

The main requirements for fuel reloads are: good reliability, minimum fuel cycle costs and flexibility of operation. Fulfilling these goals requires a background of experience. The approach to the acquisition of this experience in the particular case of BN has included over the last 15 years a proper development and cross-checking of the design methods and criteria, a continuous updating of the drawings and specifications and the qualification of adequate fabrication plants. This approach can best be outlined on the basis of the gradual implementation of the modern features of the LWR fuel. The first fuel clad with stainless steel was loaded in the BR 3 (11 MWe) in 1969 and later on (since 1974) in the SENA plant (310 MWe). Similarly, Zircaloy 4 cladding was first introduced in a reactor reload in 1969 as autoclaved cladding and later on (in 1971) the autoclaving was suppressed for the further reloads. Zircaloy 2 was loaded in DODEWAARD (51.5 MWe) in 1970. The first demonstration assembly in a PWR was a Pu-island assembly loaded in the BR 3 in 1963. It was followed by an all-Pu assembly in the same reactor in 1965 and by the loading of Pu fuels in four prototype assemblies in GARIGLIANO (160 MWe) in 1968. A full reload incorporating Pu fuel has been experienced by the supply of fuel for GARIGLIANO (BOL: 1975) and for BR 3 (BOL: 1972 and 1976). While in the early sixties the brazed design was still being utilized, the first assembly incorporating grids with springs was introduced in BR 3 in 1963. The first Inconel grids were loaded in the same reactor in 1969 and the first Zircaloy grids in 1972 (the first Zr grid has been loaded in a BWR in 1973). The experience covered successively the shrouded design (BOL: 1963), the shroudless design (BOL: 1969), a BWR assembly (BOL: 1971), a typical RCC assembly first with large diameter fuel rods (1972) and later on with small diameter fuel rods (1974). The experience on the reactivity control covered successively diluted

50 CFR 23.18 - What CITES documents are required to export Appendix-I wildlife?

Science.gov (United States)

2010-10-01

... 50 Wildlife and Fisheries 6 2010-10-01 2010-10-01 false What CITES documents are required to export Appendix-I wildlife? 23.18 Section 23.18 Wildlife and Fisheries UNITED STATES FISH AND WILDLIFE..., EXPORTATION, AND IMPORTATION OF WILDLIFE AND PLANTS (CONTINUED) CONVENTION ON INTERNATIONAL TRADE IN...

1 CFR 5.2 - Documents required to be filed for public inspection and published.

Science.gov (United States)

2010-01-01

... inspection and published. 5.2 Section 5.2 General Provisions ADMINISTRATIVE COMMITTEE OF THE FEDERAL REGISTER THE FEDERAL REGISTER GENERAL § 5.2 Documents required to be filed for public inspection and published... Register and published in the Federal Register: (a) Presidential proclamations and Executive orders in the...

HALE UAS Command and Control Communications: Step 1 - Functional Requirements Document. Version 4.0

Science.gov (United States)

2006-01-01

The High Altitude Long Endurance (HALE) unmanned aircraft system (UAS) communicates with an off-board pilot-in-command in all flight phases via the C2 data link, making it a critical component for the UA to fly in the NAS safely and routinely. This is a new requirement in current FAA communications planning and monitoring processes. This document provides a set of comprehensive C2 communications functional requirements and performance guidelines to help facilitate the future FAA certification process for civil UAS to operate in the NAS. The objective of the guidelines is to provide the ability to validate the functional requirements and in future be used to develop performance-level requirements.

A Stochastic LWR Model with Consideration of the Driver's Individual Property

International Nuclear Information System (INIS)

Tang Tieqiao; Wang Yunpeng; Yu Guizhen; Huang Haijun

2012-01-01

In this paper, we develop a stochastic LWR model based on the influences of the driver's individual property on his/her perceived density and speed deviation. The numerical results show that the driver's individual property has great effects on traffic flow only when the initial density is moderate, i.e., at this time, oscillating traffic flow will occur and the oscillating phenomena in the traffic system consisting of the conservative and aggressive drivers is more serious than that in the traffic system consisting of the conservative (aggressive) drivers.

Fracture toughness evaluation of select advanced replacement alloys for LWR core internals

Energy Technology Data Exchange (ETDEWEB)

Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chen, Xiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Fracture toughness is one of the key engineering properties required for core internal materials. Together with other properties, which are being examined such as high-temperature steam oxidation resistance, radiation hardening, and irradiation-assisted stress corrosion cracking resistance, the alloys will be down-selected for neutron irradiation study and comprehensive post-irradiation examinations. According to the candidate alloys selected under the ARRM program, ductile fracture toughness of eight alloys was evaluated at room temperature and the LWR-relevant temperatures. The tested alloys include two ferritic alloys (Grade 92 and an oxide-dispersion-strengthened alloy 14YWT), two austenitic stainless steels (316L and 310), four Ni-base superalloys (718A, 725, 690, and X750). Alloy 316L and X750 are included as reference alloys for low- and high-strength alloys, respectively. Compact tension specimens in 0.25T and 0.2T were machined from the alloys in the T-L and R-L orientations according to the product forms of the alloys. This report summarizes the final results of the specimens tested and analyzed per ASTM Standard E1820. Unlike the

Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

Energy Technology Data Exchange (ETDEWEB)

Lori Braase

2013-01-01

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and

Decay heat and gamma dose-rate prediction capability in spent LWR fuel

International Nuclear Information System (INIS)

Neely, G.J.; Schmittroth, F.

1982-08-01

The ORIGEN2 code was established as a valid means to predict decay heat from LWR spent fuel assemblies for decay times up to 10,000 year. Calculational uncertainties ranged from 8.6% to a maximum of 16% at 2.5 years and 300 years cooling time, respectively. The calculational uncertainties at 2.5 years cooling time are supported by experiment. Major sources of uncertainty at the 2.5 year cooling time were identifed as irradiation history (5.7%) and nuclear data together with calculational methods (6.3%). The QAD shielding code was established as a valid means to predict interior and exterior gamma dose rates of spent LWR fuel assemblies. A calculational/measurement comparison was done on two assemblies with different irradiation histories and supports a 35% calculational uncertainty at the 1.8 and 3.0 year decay times studied. Uncertainties at longer times are expected to increase, but not significantly, due to an increased contribution from the actinides whose inventories are assigned a higher uncertainty. The uncertainty in decay heat rises to a maximum of 16% due to actinide uncertainties. A previous study was made of the neutron emission rate from a typical Turkey Point Unit 3, Region 4 spent fuel assembly at 5 years decay time. A conservative estimate of the neutron dose rate at the assembly surface was less than 0.5 rem/hr

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Core design of super LWR with double tube water rods

International Nuclear Information System (INIS)

Wu, Jianhui; Oka, Yoshiaki

2014-01-01

Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

Short Communication on “In-situ TEM ion irradiation investigations on U{sub 3}Si{sub 2} at LWR temperatures”

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin, E-mail: ymiao@anl.gov [Argonne National Laboratory, Lemont, IL 60439 (United States); Harp, Jason [Idaho National Laboratory, Idaho Fall, ID 83415 (United States); Mo, Kun [Argonne National Laboratory, Lemont, IL 60439 (United States); Bhattacharya, Sumit [Northwestern University, Evanston, IL 60208 (United States); Baldo, Peter; Yacout, Abdellatif M. [Argonne National Laboratory, Lemont, IL 60439 (United States)

2017-02-15

The radiation-induced amorphization of U{sub 3}Si{sub 2} was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U{sub 3}Si{sub 2} specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 10{sup 15} ions/cm{sup 2} to examine their amorphization behavior under light water reactor (LWR) conditions. U{sub 3}Si{sub 2} remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

Documentation: Records and Reports.

Science.gov (United States)

Akers, Michael J

2017-01-01

This article deals with documentation to include the beginning of documentation, the requirements of Good Manufacturing Practice reports and records, and the steps that can be taken to minimize Good Manufacturing Practice documentation problems. It is important to remember that documentation for 503a compounding involves the Formulation Record, Compounding Record, Standard Operating Procedures, Safety Data Sheets, etc. For 503b outsourcing facilities, compliance with Current Good Manufacturing Practices is required, so this article is applicable to them. For 503a pharmacies, one can see the development and modification of Good Manufacturing Practice and even observe changes as they are occurring in 503a documentation requirements and anticipate that changes will probably continue to occur. Copyright© by International Journal of Pharmaceutical Compounding, Inc.

Criteria Document for B-plant's Surveillance and Maintenance Phase Safety Basis Document

International Nuclear Information System (INIS)

SCHWEHR, B.A.

1999-01-01

This document is required by the Project Hanford Managing Contractor (PHMC) procedure, HNF-PRO-705, Safety Basis Planning, Documentation, Review, and Approval. This document specifies the criteria that shall be in the B Plant surveillance and maintenance phase safety basis in order to obtain approval of the DOE-RL. This CD describes the criteria to be addressed in the S and M Phase safety basis for the deactivated Waste Fractionization Facility (B Plant) on the Hanford Site in Washington state. This criteria document describes: the document type and format that will be used for the S and M Phase safety basis, the requirements documents that will be invoked for the document development, the deactivated condition of the B Plant facility, and the scope of issues to be addressed in the S and M Phase safety basis document

Draft report: a selection methodology for LWR safety R and D programs and proposals

International Nuclear Information System (INIS)

Husseiny, A.A.; Ritzman, R.L.

1980-03-01

The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application

Draft report: a selection methodology for LWR safety R and D programs and proposals

Energy Technology Data Exchange (ETDEWEB)

Husseiny, A. A.; Ritzman, R. L.

1980-03-01

The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application.

LWR risk management by safety R and D

International Nuclear Information System (INIS)

El-Sheikh, K.A.; Damon, D.R.; Temme, M.I.

1982-01-01

This paper presents a methodology which has been developed for selecting LWR safety RandD projects. The methodology provides ranking of the RandD projects and the RandD budget allocation which minimizes public risk. The methodology contains procedures to identify institutional, organizational, legal, and contractual factors which affect the probabilities of success and use of RandD projects so that these factors can be evaluated and possibly managed.The methodology also contains a nonlinear optimization code to provide the optimum selection of RandD projects and evaluate the sensitivity of this selection to uncertainity in the input data. Application of the methodology to a test case has shown that: 1) commonly used schemes for ranking RandD projects do not necessarily lead to the optimum selection, and 2) the optimum selection is not necessarily strongly sensitive to uncertainty in the input data

Progress in Development of I2S-LWR Concept

International Nuclear Information System (INIS)

Petrovic, Bojan

2014-01-01

The paper will present the progress in developing the Integral Inherently Safe Light Water Reactor (12S-LWR) concept. This new concept aims to combine the competitive economics of a large nuclear power plant, with enhanced safety achieved by the integral primary circuit configuration (previously considered only for PWRs with power levels not exceeding several hundred MWc), and with enhanced accident tolerance (to address concerns after the Fukushima Dai-lchi accidents). Several new technologies are being developed to enable this concept, including novel silicide fuel and micro-channel primary heat exchangers. This project is performed by a multi-disciplinary multi-organization team led by Georgia Tech, including academia, a national laboratory, nuclear industry, and a power utility, wit expected participation of the University of Zagreb. (author)

Summary of aerosol code-comparison results for LWR aerosol containment tests LA1, LA2, and LA3

International Nuclear Information System (INIS)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

1987-01-01

The light-water reactor (LWR) aerosol containment experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities for the LACE tests are being coordinated at the Oak Ridge National Laboratory. For each of the six experiments, pretest calculations (for code-to-code comparisons) and blind post-test calculations (for code-to-test data comparisons) are being performed. This paper presents a summary of the pretest aerosol-code results for tests LA1, LA2, and LA3

Applications for electronic documents

International Nuclear Information System (INIS)

Beitel, G.A.

1995-01-01

This paper discusses the application of electronic media to documents, specifically Safety Analysis Reports (SARs), prepared for Environmental Restoration and Waste Management (ER ampersand WM) programs being conducted for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). Efforts are underway to upgrade our document system using electronic format. To satisfy external requirements (DOE, State, and Federal), ER ampersand WM programs generate a complement of internal requirements documents including a SAR and Technical Safety Requirements along with procedures and training materials. Of interest, is the volume of information and the difficulty in handling it. A recently prepared ER ampersand WM SAR consists of 1,000 pages of text and graphics; supporting references add 10,000 pages. Other programmatic requirements documents consist of an estimated 5,000 pages plus references

Utility guidance to advanced LWR designers

International Nuclear Information System (INIS)

Yedidia, J.M.

1990-01-01

The purpose of this paper is to describe the process envisioned for the development of advanced reactors for future use by the utility industry. The role of the potential utility customer is gradually evolving from that of an owner-operator of such plants to that of a sponsor-participant in the actual design process. The author discusses development of a set of utility requirements, intended to describe in detail utility needs and expectations relative to the performance of future reactors. The reactor vendors, who participated actively in the preparation of the requirements documents, pledged to make every effort to meet them in their future designs. At that stage, when the requirements have been finalized and agreed to by all parties involved, including the Nuclear Regulatory Commission, the utilities were expected to move to the sidelines and wait for the reactor vendors to come up with the product

Trends in U.S. nuclear standards development

International Nuclear Information System (INIS)

Crowley, J.H.; Kaminski, R.S.

1987-01-01

Regulation of the U.S Nuclear Power industry has been extensive during the 1970's. Key to this situation has been the evolution in the 'interpretation' of the rules, regulations and consensus standards which have been incorporated into NRC guidance documents by reference or endorsement. The resulting increase in the number and complexity of LWR construction requirements has significantly increased the labor content of an LWR construction project. The authors believe that existing nuclear related consensus standards should be reviewed with the objective of modifying the standards to improve the efficiency and productivity of engineering, craft, and non-manual personnel. (author)

European Utility Requirements: leveling the European electricity producers' playing ground for new NPPs

International Nuclear Information System (INIS)

Bernard Roche

2006-01-01

Full text of publication follows: Since 1992, the European Utility Requirement (EUR) document has been developed by the major European electricity producers. The main driver to this work has been the construction of a unified European market. The electricity producers have set out design requirements adapted to this new European environment, while keeping in mind experience feedback from operating NPPs worldwide. The EUR document is now fully operational and its set of generic requirements have been recently used as bid specification in Finland and in China. The EUR document keeps developing in two directions: 1- completing the assessment of the projects that could be proposed by the vendors for the European market. Five projects have been assessed between 1999 and 2002: BWR90, EPR, EP1000, ABWR and SWR1000. Two new projects are being assessed, the Westinghouse AP1000 and the Russian VVER AES92. It is currently planned to publish these two new assessments in the first half of 2006. Others may be undertaken meanwhile. 2- revision of the generic requirements. A revision C of the volume 4 dedicated to power generation plant is being completed. It includes responses to vendors comments and feedback from the TVO call for bid for Finland 5. A revision D of the volumes 1 and 2 dedicated to nuclear islands is foreseen. The main contributions to this revision are the harmonization actions going on in Europe about nuclear safety (WENRA study on reactor safety harmonization, EC works, evolution of the IAEA guides and requirements), the harmonization works on the conditions of connection to the European HV grid as well as harmonization works on other matters, like codes and standards. This has given a unified frame in which the future nuclear plants can be designed and built. In this frame development of standards designs usable throughout Europe without major design change is possible, thus helping to increase competition, and ultimately to save investment and operating costs

Safeguards Guidance Document for Designers of Commercial Nuclear Facilities: International Nuclear Safeguards Requirements and Practices For Uranium Enrichment Plants

Energy Technology Data Exchange (ETDEWEB)

Robert Bean; Casey Durst

2009-10-01

This report is the second in a series of guidelines on international safeguards requirements and practices, prepared expressly for the designers of nuclear facilities. The first document in this series is the description of generic international nuclear safeguards requirements pertaining to all types of facilities. These requirements should be understood and considered at the earliest stages of facility design as part of a new process called “Safeguards-by-Design.” This will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards verification activities. The following summarizes the requirements for international nuclear safeguards implementation at enrichment plants, prepared under the Safeguards by Design project, and funded by the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Office of NA-243. The purpose of this is to provide designers of nuclear facilities around the world with a simplified set of design requirements and the most common practices for meeting them. The foundation for these requirements is the international safeguards agreement between the country and the International Atomic Energy Agency (IAEA), pursuant to the Treaty on the Non-proliferation of Nuclear Weapons (NPT). Relevant safeguards requirements are also cited from the Safeguards Criteria for inspecting enrichment plants, found in the IAEA Safeguards Manual, Part SMC-8. IAEA definitions and terms are based on the IAEA Safeguards Glossary, published in 2002. The most current specification for safeguards measurement accuracy is found in the IAEA document STR-327, “International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials,” published in 2001. For this guide to be easier for the designer to use, the requirements have been restated in plainer language per expert interpretation using the source documents noted. The safeguards agreement is fundamentally a

Computer program of iodine removal in the LWR containment vessel under LOCA conditions, MIRA-PB

International Nuclear Information System (INIS)

Nishio, Gunji; Tanaka, Mitsugu; Tamura, Tomohiko.

1978-03-01

LWR plants have a containment system for reactor safety consisting of spray and air cleaning filter. R.L.Ritzman of Battele Columbus Lab. developed computer code MIRAP/MIRAB for predicting iodine removal by containment system for PWR and BWR; which has some problem, however. The computer code MIRA-PB prepared by the authors is a modification of MIRAP/MIRAB. (auth.)

Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2E, Physical descriptions of LWR nonfuel assembly hardware, Appendix 2F, User's guide to the LWR nonfuel assembly data base

International Nuclear Information System (INIS)

1987-12-01

This appendix includes a two to three page Physical Description report for each Non-fuel Assembly (NFA) Hardware item identified from the current data. Information was obtained via subcontracts with these NFA hardware vendors: Babcock and Wildox, Combustion Engineering and Westinghouse. Data for some NFA hardware are not available. For such hardware, the information shown in this report was obtained from the open literature. Efforts to obtain additional information are continuing. NFA hardware can be grouped into six categories: BWR Channels, Control Elements, Guide Tube Plugs/Orifice Rods, Instrumentation, Neutron Poisons, and Neutron Sources. This appendix lists Physical Description reports alphabetically by vendor within each category. Individual Physical Description reports can be generated interactively through the menu-driven LWR Non-Fuel Assembly Hardware Data Base system. These reports can be viewed on the screen, directed to a printer, or saved in a text file for later use. Special reports and compilations of specific data items can be produced on request

Validation of the LWR-EIR methods for the evaluation of compact beds

International Nuclear Information System (INIS)

Foskolos, K.; Grimm, P.; Maeder, C.; Paratte, J.M.

1983-10-01

The EIR code system for the calculation of light water reactors is presented and the methods used are briefly described. The application of the system on various types of critical experiments and benchmark problems proves its good precision, even for heterogeneous configurations with strong neutron absorbers like Boral. As the accuracy of the multiplication factor ksub(eff) is always better than 0.5% for normal LWR configurations, this code system is validated for the calculation of such configurations with a safety margin of 1.5% on ksub(eff). (Auth.)

The control of documents maintained in the electronic media and the requirements of NBR ISO/IEC 17025 and adequacy for PNQ

Directory of Open Access Journals (Sweden)

Anselmo Ferreira de Castro

2006-01-01

Full Text Available During the assessments carried out by the authors at the diverse laboratories pertaining to the network of accredited laboratories - known as the Brazilian Calibration Network (RBC - the authors verified the difficulties that these laboratories had in meeting the requirements of NBR ISO/IEC 17025:2001 (Standard, related to the maintenance of documents in the electronic media. More and more laboratories are substituting traditional control with electronic document control, which allows for more agility in the recovery of information. These laboratories implement policies and procedures; however, they still feel insecure, in some way, as to how to meet all of these requirements, thereby giving rise to difficulties in the implementation of such. The aim of this paper is to discuss the control of documents stored in the electronic media, adopting requirement 4.3 of the Standard as the reference, and aiming, in this manner, to harmonize the assessment process of the procedures of document management in the electronic media, and to assist the laboratories in the interpretation of the Standard, so that they may implement systems that are adequate to their actual necessities and to their structural size, while at the same time complying with the referred to requirement. This work will not broach the treatment given to the records (requirement 4.12 of the Standard, facts (requirement 5.4.7 nor to the electronic transmission of results (requirement 5.10.7, leaving these subjects for posterior discussions.

Conceptual development of a complete LWR reload design methodology based on generalized perturbation theory

International Nuclear Information System (INIS)

White, J.R.

1986-01-01

A new approach for the physics design and analysis of LWR reload cores is developed and demonstrated through several practical applications. The new design philosophy uses first- and second-order response derivatives to predict the important reactor performance characteristics (power peaking, reactivity coefficients, etc.) for any number of possible material configurations (assembly shuffling and burnable poison loadings). The response derivatives are computed using generalized perturbation theory (GPT) techniques. This report describes in detail an idealized GPT-based design system. The idealized system would contain individual modules to generate the required first-order and higher-order sensitivity data. It would also contain at least two major application codes; one for core design optimization and the other for evaluation of several safety parameters of interest in off-normal situations. This ideal system would be fully automated, user-friendly, and quite flexible in its ability to provide a variety of design and analysis capabilities. Information gained form these three studies gives a good foundation for the development of a complete integrated design package

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 2

International Nuclear Information System (INIS)

1994-04-01

The Quality Assurance Functional Area Requirements Identification Document (RID), addresses the programmatic requirements that ensure risks and environmental impacts are minimized, ensure safety, reliability, and performance are maximized through the application of effective management systems commensurate with the risks posed by the Tank Farm Facility and its operation. This RID incorporates guidance intended to provide Tank Farms management with the necessary requirements information to develop, upgrade, or assess the effectiveness of a Quality Assurance Program in the performance of organizational and functional activities. Quality Assurance is defined as all those planned and systematic actions necessary to provide adequate confidence that a facility, structure, system, or component will perform satisfactorily and safely in service. This document will provide the specific requirements to meet DNFSB recommendations and the guidance provided in DOE Order 5700.6C, utilizing industry codes, standards, regulatory guidelines, and industry good practices that have proven to be essential elements for an effective and efficient Quality Assurance Program as the nuclear industry has matured over the last thirty years

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 2

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

The Quality Assurance Functional Area Requirements Identification Document (RID), addresses the programmatic requirements that ensure risks and environmental impacts are minimized, ensure safety, reliability, and performance are maximized through the application of effective management systems commensurate with the risks posed by the Tank Farm Facility and its operation. This RID incorporates guidance intended to provide Tank Farms management with the necessary requirements information to develop, upgrade, or assess the effectiveness of a Quality Assurance Program in the performance of organizational and functional activities. Quality Assurance is defined as all those planned and systematic actions necessary to provide adequate confidence that a facility, structure, system, or component will perform satisfactorily and safely in service. This document will provide the specific requirements to meet DNFSB recommendations and the guidance provided in DOE Order 5700.6C, utilizing industry codes, standards, regulatory guidelines, and industry good practices that have proven to be essential elements for an effective and efficient Quality Assurance Program as the nuclear industry has matured over the last thirty years.

Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

International Nuclear Information System (INIS)

Chopra, O.K.; Shack, W.J.

1998-03-01

The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented

H2FIRST Hydrogen Contaminant Detector Task: Requirements Document and Market Survey

Energy Technology Data Exchange (ETDEWEB)

Terlip, Danny [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Ainscough, Chris [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Buttner, William [National Renewable Energy Laboratory (NREL), Golden, CO (United States); McWhorter, Scott [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-04-20

The rollout of hydrogen fueling stations, and the fuel cell electric vehicles (FCEV) they support, requires the assurance of high quality hydrogen at the dispensing point. Automotive fuel cells are sensitive to a number of chemicals that can be introduced into the dispensed fuel at multiple points. Quality assurance and quality control methods are employed by the industry to ensure product quality, but they are not completely comprehensive and can fail at various points in the hydrogen pathway from production to dispensing. This reality leaves open the possibility of a station unknowingly dispensing harmful contaminants to a FCEV which, depending on the contaminant, may not be discovered until the FCEV is irreparably damaged. This situation is unacceptable. A hydrogen contaminant detector (HCD), defined as a combination of a gas analyzer and the components necessary for fuel stream integration, installed at hydrogen stations is one method for preventing poor quality gas from reaching an FCEV. This document identifies the characteristics required of such a device by industry and compares those requirements with the current state of commercially available gas analysis technology.

Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

International Nuclear Information System (INIS)

Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

1977-01-01

The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

Low leaching and low LWR photoresist development for 193 nm immersion lithography

Science.gov (United States)

Ando, Nobuo; Lee, Youngjoon; Miyagawa, Takayuki; Edamatsu, Kunishige; Takemoto, Ichiki; Yamamoto, Satoshi; Tsuchida, Yoshinobu; Yamamoto, Keiko; Konishi, Shinji; Nakano, Katsushi; Tomoharu, Fujiwara

2006-03-01

With no apparent showstopper in sight, the adoption of ArF immersion technology into device mass production is not a matter of 'if' but a matter of 'when'. As the technology matures at an unprecedented speed, many of initial technical difficulties have been cleared away and the use of a protective layer known as top coat, initially regarded as a must, now becomes optional, for example. Our focus of interest has also sifted to more practical and production related issues such as defect reducing and performance enhancement. Two major types of immersion specific defects, bubbles and a large number of microbridges, were observed and reported elsewhere. The bubble defects seem to decrease by improvement of exposure tool. But the other type defect - probably from residual water spots - is still a problem. We suspect that the acid leaching from resist film causes microbridges. When small water spots were remained on resist surface after exposure, acid catalyst in resist film is leaching into the water spots even though at room temperature. After water from the spot is dried up, acid molecules are condensed at resist film surface. As a result, in the bulk of resist film, acid depletion region is generated underneath the water spot. Acid catalyzed deprotection reaction is not completed at this acid shortage region later in the PEB process resulting in microbridge type defect formation. Similar mechanism was suggested by Kanna et al, they suggested the water evaporation on PEB plate. This hypothesis led us to focus on reducing acid leaching to decrease residual water spot-related defect. This paper reports our leaching measurement results and low leaching photoresist materials satisfying the current leaching requirements outlined by tool makers without topcoat layer. On the other hand, Nakano et al reported that the higher receding contact angle reduced defectivity. The higher receding contact angle is also a key item to increase scan speed. The effort to increase the

Requirements Document for Development of a Livermore Tomography Tools Interface

Energy Technology Data Exchange (ETDEWEB)

Seetho, I. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-02-09

In this document, we outline an exercise performed at LLNL to evaluate the user interface deficits of a LLNL-developed CT reconstruction software package, Livermore Tomography Tools (LTT). We observe that a difficult-to-use command line interface and the lack of support functions compound to generate a bottleneck in the CT reconstruction process when input parameters to key functions are not well known. Through the exercise of systems engineering best practices, we generate key performance parameters for a LTT interface refresh, and specify a combination of back-end (“test-mode” functions) and front-end (graphical user interface visualization and command scripting tools) solutions to LTT’s poor user interface that aim to mitigate issues and lower costs associated with CT reconstruction using LTT. Key functional and non-functional requirements and risk mitigation strategies for the solution are outlined and discussed.

Survey of LWR environmental control technology performance and cost

International Nuclear Information System (INIS)

Heeb, C.M.; Aaberg, R.L.; Cole, B.M.; Engel, R.L.; Kennedy, W.E. Jr.; Lewallen, M.A.

1980-03-01

This study attempts to establish a ranking for species that are routinely released to the environment for a projected nuclear power growth scenario. Unlike comparisons made to existing standards, which are subject to frequent revision, the ranking of releases can be used to form a more logical basis for identifying the areas where further development of control technology could be required. This report describes projections of releases for several fuel cycle scenarios, identifies areas where alternative control technologies may be implemented, and discusses the available alternative control technologies. The release factors were used in a computer code system called ENFORM, which calculates the annual release of any species from any part of the LWR nuclear fuel cycle given a projection of installed nuclear generation capacity. This survey of fuel cycle releases was performed for three reprocessing scenarios (stowaway, reprocessing without recycle of Pu and reprocessing with full recycle of U and Pu) for a 100-year period beginning in 1977. The radioactivity releases were ranked on the basis of a relative ranking factor. The relative ranking factor is based on the 100-year summation of the 50-year population dose commitment from an annual release of radioactive effluents. The nonradioactive releases were ranked on the basis of dilution factor. The twenty highest ranking radioactive releases were identified and each of these was analyzed in terms of the basis for calculating the release and a description of the currently employed control method. Alternative control technology is then discussed, along with the available capital and operating cost figures for alternative control methods

Performing the processing required for automatically get a PDF/A version of the CERN Library documentation

CERN Document Server

Molina Garcia-Retamero, Antonio

2015-01-01

The aim of the project was to perform the processing required for automatically get a PDF/A version of the CERN Library documentation. For this, it is necessary to extract as much metadata as possible from the sources files, inject the required data into the original source files creating new ones ready for being compiled with all related dependencies. Besides, I’ve proposed the creation of a HTML version consistent with the PDF and navigable for easy access, I’ve been trying to perform some Natural Language Processing for extracting metadata, I’ve proposed the injection of the cern library documentation into the HTML version of the long writeups where it is referenced (for instance, when a CERN Library function is referenced in a sample code) Finally, I’ve designed and implemented a Graphical User Interface in order to simplify the process for the user.

Joint Simulation System (JSIMS) Functional Requirements Document (FRD); A User's Perspective on the Future, Version 1.O

National Research Council Canada - National Science Library

1996-01-01

The purpose of this document is to define JSIMS functional requirements in a level of detail that is meaningful to both the JSIMS developmental community and the expected future users of the system...

IDC System Specification Document.

Energy Technology Data Exchange (ETDEWEB)

Clifford, David J.

2014-12-01

This document contains the system specifications derived to satisfy the system requirements found in the IDC System Requirements Document for the IDC Reengineering Phase 2 project. Revisions Version Date Author/Team Revision Description Authorized by V1.0 12/2014 IDC Reengineering Project Team Initial delivery M. Harris

Fracture probability evaluation of a LWR pressure vessel

International Nuclear Information System (INIS)

Grandemange, J.; Pellissier-Tanon, A.; Quero, J.; Carnino, A.; Dufresne, J.

1978-01-01

Fracture probability evaluation, of a LWR pressure vessel have been performed in the past, using statistical data from conventional plant. A more accurate evaluation has been requested in 1976 from the SCSIN to the CEA. With this object, a joint collaboration agreement has been signed between CEA, EURATOM/ISPRA and FRAMATOME. The whole program proceeding from this agreement is managed by a joint board including the three partners. The basic objective of this program is to develop a method which integrates, or makes it possible to integrate at a later stage, the greatest number of significant parameters. Also, in order to prepare the practical applications, a special effort is being made to collect the data corresponding to these parameters. Parallel basic research program have been launched in order to clarify our knowledge on some important parts of the main factors contributing to the evaluation. The results of this research will be progressively introduced into the method or will help checking its validity

Convergence studies of deterministic methods for LWR explicit reflector methodology

International Nuclear Information System (INIS)

Canepa, S.; Hursin, M.; Ferroukhi, H.; Pautz, A.

2013-01-01

The standard approach in modem 3-D core simulators, employed either for steady-state or transient simulations, is to use Albedo coefficients or explicit reflectors at the core axial and radial boundaries. In the latter approach, few-group homogenized nuclear data are a priori produced with lattice transport codes using 2-D reflector models. Recently, the explicit reflector methodology of the deterministic CASMO-4/SIMULATE-3 code system was identified to potentially constitute one of the main sources of errors for core analyses of the Swiss operating LWRs, which are all belonging to GII design. Considering that some of the new GIII designs will rely on very different reflector concepts, a review and assessment of the reflector methodology for various LWR designs appeared as relevant. Therefore, the purpose of this paper is to first recall the concepts of the explicit reflector modelling approach as employed by CASMO/SIMULATE. Then, for selected reflector configurations representative of both GII and GUI designs, a benchmarking of the few-group nuclear data produced with the deterministic lattice code CASMO-4 and its successor CASMO-5, is conducted. On this basis, a convergence study with regards to geometrical requirements when using deterministic methods with 2-D homogenous models is conducted and the effect on the downstream 3-D core analysis accuracy is evaluated for a typical GII deflector design in order to assess the results against available plant measurements. (authors)

Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations

Energy Technology Data Exchange (ETDEWEB)

Brown, S. J. [ed.

1980-04-01

The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation.

ELECTRICAL POWER SYSTEM DESCRIPTION DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

M. Maniyar

2004-06-22

The purpose of this revision of the System Description Document (SDD) is to establish requirements that drive the design of the electrical power system and their bases to allow the design effort to proceed to License Application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience are design engineers. This type of SDD leads and follows the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential to performing the design process. This SDD follows the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. Functional and operational requirements applicable to this system are obtained from ''Project Functional and Operational Requirements'' (F&OR) (Siddoway, 2003). Other requirements to support the design process have been taken from higher level requirements documents such as ''Project Design Criteria Document'' (PDC) (Doraswamy 2004), the fire hazards analyses, and the preclosure safety analysis. The above mentioned low-level documents address ''Project Requirements Document'' (PRD) (Canori and Leitner 2003) requirements. This SDD includes several appendices with supporting information. Appendix B lists key system charts, diagrams, drawings, and lists; and Appendix C is a list of system procedures.

Supplemental design requirements document enhanced radioactive and mixed waste storage Phase V Project W-112

International Nuclear Information System (INIS)

Ocampo, V.P.; Boothe, G.F.; Greager, T.M.; Johnson, K.D.; Kooiker, S.L.; Martin, J.D.

1994-11-01

This document provides additional and supplemental information to WHC-SD-W112-FDC-001, Project W-112 for radioactive and mixed waste storage. It provides additional requirements for the design and summarizes Westinghouse Hanford Company key design guidance and establishes the technical baseline agreements to be used for definitive design of the Project W-112 facilities

Effect of Casmo-5 cross-section data and doppler temperature definitions on LWR reactivity initiated accidents - 166

International Nuclear Information System (INIS)

Grandi, G.; Smith, K.; Xu, Z.; Rhodes, J.

2010-01-01

During LWR Reactivity Initiated Accidents (RIA), the accurate evaluation of the Doppler reactivity feedback depends on the Doppler coefficient computed by the lattice physics code (e.g. CASMO-5), and on the effective Doppler temperature computed by the transient code (e.g. SIMULATE-3K) using the non-uniform intra-pellet temperature profile. CASMO-5 has many new features compared with its predecessor. Among them, the replacement of the L-library (based primarily on ENDF/B IV data) by the latest available nuclear data (ENDF/B VII.0), and the Monte Carlo based resonance elastic scattering model to overcome deficiencies in NJOY modeling have a significant impact on the fuel temperature coefficient, and hence on LWR RIA. The Doppler temperature effect in thermal reactors is driven by the 238 U absorption. The different effective Doppler temperature definitions, available in the literature, try to capture the considerable self-shielding of the 238 U absorption that occurs in the pellet surface by defining an appropriate fuel temperature to compute cross-sections. In this work, we investigate the effect of the nuclear data generated by CASMO-5 on RIA, as well as the impact of different effective Doppler temperature definitions, including one proposed by the authors. It is concluded: 1) LWR RIA evaluated using CASMO-5 cross section data will be milder because the energy released is ∼10% smaller; 2) the prompt enthalpy rise is barely affected by the choice of the Doppler temperature definition; and 3) the peak fuel enthalpy is affected by the choice of the Doppler temperature definition, the under-prediction of the Doppler reactivity by the 'NEA' Doppler temperature results in a conservative estimate of the peak fuel enthalpy. (authors)

Results of LWR core transient benchmarks

International Nuclear Information System (INIS)

Finnemann, H.; Bauer, H.; Galati, A.; Martinelli, R.

1993-10-01

LWR core transient (LWRCT) benchmarks, based on well defined problems with a complete set of input data, are used to assess the discrepancies between three-dimensional space-time kinetics codes in transient calculations. The PWR problem chosen is the ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal - hydraulic models of the codes. The 63 sets of submitted solutions are analyzed by comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations. The BWR problems considered are reactivity excursions caused by cold water injection and pressurization events. In the present paper, only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution the evaluation of the 8 sets of BWR contributions relies on a synthetic comparative discussion. The results of this first phase of LWRCT benchmark calculations are quite satisfactory, though there remain some unresolved issues. It is therefore concluded that even more challenging problems can be successfully tackled in a suggested second test phase. (authors). 46 figs., 21 tabs., 3 refs

Review of tellurium release rates from LWR fuel elements under accident conditions

International Nuclear Information System (INIS)

Lorenz, R.A.; Beahm, E.C.; Wichner, R.P.

1983-01-01

Although fission product tellurium presents a potentially significant radiohazard, its release and transport in source-term experiments is frequently overlooked because it does not possess a readily measurable, gamma emission; moreover, a recent study emphasized noble gas, iodine and cesium release from LWR fuel elements because of the large data base that exists for these materials. Some new tests show that in some cases tellurium may be held up in core material to a greater degree than previously assumed - an observation that prompts a careful reappraisal of the existing tellurium-release data and its chemical foundation

Identification of the impacts of maintenance and testing upon the safety of LWR power plants. Final report

International Nuclear Information System (INIS)

Husseiny, A.A.; Sabri, Z.A.; Turnage, J.J.

1980-04-01

The present study was designed to identify the impact of maintenance and testing (M and T) upon the safety of LWR power plants. The study involved data extraction from various sources reporting safety-related and operation-related nuclear power plant experience. Primary sources reviewed, including Licensee Event Reports (LER's) submitted to the NRC, revealed that only ten percent of events reported could be identified as M and T problems. The collected data were collated in a manner that would allow identification of principal types of problems which are associated with the performance of M and T tasks in LWR power plants. Frequencies of occurrence of events and their general endemic nature were analyzed using data clustering and pattern recognition techniques, as well as chi-square analyses for sparse contingency tables. The results of these analyses identified seven major categories of M and T error modes which were related to individual facilities and reactor type. Data review indicated that few M and T problems were directly related to procedural inadequacies, with the majority of events being attributable to human error

Proposal and analysis of the benchmark problem suite for reactor physics study of LWR next generation fuels

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-10-01

In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has established the Working Party on Reactor Physics for LWR Next Generation Fuels. The next generation fuels mean the ones aiming for further extended burn-up such as 70 GWd/t over the current design. The Working Party has proposed six benchmark problems, which consists of pin-cell, PWR fuel assembly and BWR fuel assembly geometries loaded with uranium and MOX fuels, respectively. The specifications of the benchmark problem neglect some of the current limitations such as 5 wt% {sup 235}U to achieve the above-mentioned target. Eleven organizations in the Working Party have carried out the analyses of the benchmark problems. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified. In this report, details of the benchmark problems, result by each organization, and their comparisons are presented. (author)

Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments

Energy Technology Data Exchange (ETDEWEB)

Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kim, Young -Jin [GE Global Research Center, Schenectady, NY (United States); Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-11-28

Nano-infiltration and transient eutectic phase (NITE) sintering was developed for fabrication of nuclear grade SiC composites. We produced monolithic SiC ceramics using NITE sintering, as candidates for accident-tolerant fuels in light-water reactors (LWRs). In this work, we exposed three different NITE chemistries (yttria-alumina [YA], ceria-zirconia-alumina [CZA], and yttria-zirconia-alumina [YZA]) to autoclave conditions simulating LWR coolant loops. The YZA was most corrosion resistant, followed by CZA, with YA being worst. High-resolution elemental analysis using scanning transmission electron microscopy (STEM) X-ray mapping combined with multivariate statistical analysis (MVSA) datamining helped explain the differences in corrosion. YA-NITE lost all Al from the corroded region and the ytttria reformed into blocky precipitates. The CZA material lost all Al from the corroded area, and the YZA – which suffered the least corrosion –retained some Al in the corroded region. Lastly, the results indicate that the YZA-NITE SiC is most resistant to hydrothermal corrosion in the LWR environment.

Experience of European LWR irradiated fuel transport: the first five hundred tonnes

International Nuclear Information System (INIS)

Curtis, H.W.

1978-01-01

The paper describes the service provided by an international company specializing in the transport of LWR irradiated fuel throughout Europe. Methods of transport used to the reprocessing plants at La Hague and Windscale include road transport of 38 te flasks over the whole route; transport of flasks between 55 and 105 te by rail, with rail-head and the reprocessing plant, where required, performed by road using heavy trailers; roll-on, roll-off sea ferries; and charter ships. Different modes of transport have been developed to cater for the various limitations on access to reactor sites arising from geographical and routing considerations. The experience of transporting more than 500 tonnes of irradiated uranium from twenty-one power reactors is used to illustrate the flexibility which the transport organization requires when the access and handling facilities are different at almost every reactor. Variations in fuel cross sections and lengths of fuel elements used in first generation reactors created the need for first generation flasks with sufficient variants to accommodate all reactor fuels but the trend now is towards standardization of flasks to perhaps two basic types. The safety record of irradiated fuel transport is examined with explanation of the means whereby this has been achieved. The problems of programming the movement of a pool of eighteen flasks for twenty-one reactors in eight countries are discussed together with the steps taken to ensure that the service operates fairly to give priority to those reactors with the greatest problems. The transport of irradiated fuel across several national frontiers is an international task which requires an international company. The transport of European irradiated fuel can be presented as an example of international collaboration which works

The LER/LWR metrology challenge for advance process control through 3D-AFM and CD-SEM

Science.gov (United States)

Faurie, P.; Foucher, J.; Foucher, A.-L.

2009-12-01

The continuous shrinkage in dimensions of microelectronic devices has reached such level, with typical gate length in advance R&D of less than 20nm combine with the introduction of new architecture (FinFET, Double gate...) and new materials (porous interconnect material, 193 immersion resist, metal gate material, high k materials...), that new process parameters have to be well understood and well monitored to guarantee sufficient production yield in a near future. Among these parameters, there are the critical dimensions (CD) associated to the sidewall angle (SWA) values, the line edge roughness (LER) and the line width roughness (LWR). Thus, a new metrology challenge has appeared recently and consists in measuring "accurately" the fabricated patterns on wafers in addition to measure the patterns on a repeatable way. Therefore, a great effort has to be done on existing techniques like CD-SEM, Scatterometry and 3D-AFM in order to develop them following the two previous criteria: Repeatability and Accuracy. In this paper, we will compare the 3D-AFM and CD-SEM techniques as a mean to measure LER and LWR on silicon and 193 resist and point out CD-SEM impact on the material during measurement. Indeed, depending on the material type, the interaction between the electron beam and the material or between the AFM tip and the material can vary a lot and subsequently can generate measurements bias. The first results tend to show that depending on CD-SEM conditions (magnification, number of acquisition frames) the final outputs can vary on a large range and therefore show that accuracy in such measurements are really not obvious to obtain. On the basis of results obtained on various materials that present standard sidewall roughness, we will show the limit of each technique and will propose different ways to improve them in order to fulfil advance roadmap requirements for the development of the next IC generation.

Recommended HSE-7 documents hierarchy

International Nuclear Information System (INIS)

Klein, R.B.; Jennrich, E.A.; Lund, D.M.; Danna, J.G.; Davis, K.D.; Rutz, A.C.

1990-01-01

This report recommends a hierarchy of waste management documents at Los Alamos National Laboratory (LANL or ''Laboratory''). The hierarchy addresses documents that are required to plan, implement, and document waste management programs at Los Alamos. These documents will enable the waste management group and the six sections contained within that group to satisfy requirements that are imposed upon them by the US Department of Energy (DOE), DOE Albuquerque Operations, US Environmental Protection Agency, various State of New Mexico agencies, and Laboratory management

Safety-related investigations on power distribution in MOX fuel elements in LWR cores

International Nuclear Information System (INIS)

Kramer, E.; Langenbuch, S.

1991-01-01

For the concept of thermal recycling various fuel assembly designs have been developped during the last years. An overview is given describing the present status of MOX-fuel assembly design for PWR and BWR. The local power distribution within the MOX-fuel assembly and influences between neighbouring MOX- and Uranium fuel assemblies have been analyzed by own calculations. These investigations are limited to specific aspects of the spatial power distribution, which are related to the use of MOX-fuel assemblies within the reactor core of LWR. (orig.) [de

Supplemental design requirements document, Multifunction Waste Tank Facility, Project W-236A. Revision 1

International Nuclear Information System (INIS)

Groth, B.D.

1995-01-01

The Multi-Function Waste Tank Facility (MWTF) consists of four, nominal 1 million gallon, underground double-shell tanks, located in the 200-East area, and two tanks of the same capacity in the 200-West area. MWTF will provide environmentally safe storage capacity for wastes generated during remediation/retrieval activities of existing waste storage tanks. This document delineates in detail the information to be used for effective implementation of the Functional Design Criteria requirements

Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

Energy Technology Data Exchange (ETDEWEB)

Schneider, K.J.; Kabele, T.J.

1979-09-01

To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

International Nuclear Information System (INIS)

Schneider, K.J.; Kabele, T.J.

1979-09-01

To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems

22 CFR 203.7 - IPVO initial documentation requirements.

Science.gov (United States)

2010-04-01

... structure; (4) Statement of tax exemption or a comparable document from the country of its origin; (5... IPVO's country of domicile by an independent certified public accountant (CPA) and in U.S. dollars; (6...

ELECTRICAL SUPPORT SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

Roy, S.

2004-01-01

The purpose of this revision of the System Design Description (SDD) is to establish requirements that drive the design of the electrical support system and their bases to allow the design effort to proceed to License Application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience/users are design engineers. This type of SDD both ''leads'' and ''trails'' the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. Functional and operational requirements applicable to electrical support systems are obtained from the ''Project Functional and Operational Requirements'' (F andOR) (Siddoway 2003). Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (Doraswamy 2004), and fire hazards analyses. The above-mentioned low-level documents address ''Project Requirements Document'' (PRD) (Canon and Leitner 2003) requirements. This SDD contains several appendices that include supporting information. Appendix B lists key system charts, diagrams, drawings, and lists, and Appendix C includes a list of system procedures

Removal of 14C-contaminated CO2 from simulated LWR fuel reprocessing off-gas by utilizing the reaction between CO2 and alkaline hydroxides in either slurry or solid form

International Nuclear Information System (INIS)

Holladay, D.W.; Haag, G.L.

1979-01-01

An important consideration in the design of a LWR fuel reprocessing plant is the removal of 14 C-contaminated CO 2 from the process off-gas. The separation and fixation of essentially all the CO 2 from the simulated off-gas can be accomplished by reaction with alkaline slurries in agitated tank-type contactors. Based on efficacy for CO 2 removal, consideration of reactant cost, and stability of the carbonate product as related to long-term storage requirements, the two most promising slurry reactants for CO 2 removal from low CO 2 -content feed gases are Ca(OH) 2 and Ba(OH) 2 . The removal of 14 C-contaminated CO 2 from simulated LWR off-gases was studied as a function of both operating conditions and varying sizes of bench-scale design. Parametrically, the effects on the CO 2 removal rate of feed composition (330 ppM - 4.47% CO 2 ), impeller speed (325 to 650 rpm), superficial velocity (5 to 80 cm/min), reactants [Mg(OH) 2 , NaOH], contactor size (20.3 cm and 27.3 cm ID), and type of operation (semibatch or continuous slurry) were deterined

NUPEC proves reliability of LWR fuel assemblies

International Nuclear Information System (INIS)

Anon.

1987-01-01

It is very important in assuring the safety of nuclear reactors to confirm the reliability of fuel assemblies. The test program of the Nuclear Power Engineering Center on the reliability of fuel assemblies has verified the high performance and reliability of Japanese LWR fuels, and confirmed the propriety of their design and fabrication. This claim is based on the data obtained from the fuel assemblies irradiated in commercial reactors. The NUPEC program includes irradiation test which has been conducted for 11 years since fiscal 1976, and the maximum thermal loading test using the out of pile test facilities simulating a real reactor which has been continued since fiscal 1978. The irradiation test on BWR fuel assemblies in No.3 reactor in Fukushima No.1 Nuclear Power Station, Tokyo Electric Power Co., Inc., and on PWR fuel assemblies in No.3 reactor in Mihama Power Station, Kansai Electric Power Co., Inc., and the maximum thermal loading test on BWR and PWR fuel assemblies are reported. The series of postirradiation examination of the fuel assemblies used for commercial reactors was conducted for the first time in Japan, and the highly systematic data on 27 items were obtained. (Kako, I.)

Development of a data bank system for LWR integral experiment

International Nuclear Information System (INIS)

Naito, Yoshitaka; Aoyagi, Hideo

1983-01-01

A data bank system for LWR integral experiment has been developed for the purpose of alleviating various efforts associated with the verification of computer codes. The final aim of this system is such that the imput data for the code to be verified can be easily obtained, and the results of calculation can be obtained in the form of the comparison with measurement. Geometry and material composition as well as measured data are stored in the data bank. This data bank system is composed of four sub-programs; (1) registration program, (2) information retrieval program, (3) maintenance program, and (4) figure representation program. In this report, the structure of this data bank system and how to use the system are explained. An example of the use of this system is also included. (Aoki, K.)

Integral Inherently Safe Light Water Reactor (I2S-LWR)

International Nuclear Information System (INIS)

Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

2017-01-01

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

Monitored Geologic Repository Project Description Document

International Nuclear Information System (INIS)

Curry, P. M.

2001-01-01

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)

Irradiation effects on thermal properties of LWR hydride fuel

Energy Technology Data Exchange (ETDEWEB)

Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

2017-04-01

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

International Nuclear Information System (INIS)

Olson, W.O.; Nigg, D.W.

1981-05-01

The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

Extremely secure identification documents

International Nuclear Information System (INIS)

Tolk, K.M.; Bell, M.

1997-09-01

The technology developed in this project uses biometric information printed on the document and public key cryptography to ensure that an adversary cannot issue identification documents to unauthorized individuals or alter existing documents to allow their use by unauthorized individuals. This process can be used to produce many types of identification documents with much higher security than any currently in use. The system is demonstrated using a security badge as an example. This project focused on the technologies requiring development in order to make the approach viable with existing badge printing and laminating technologies. By far the most difficult was the image processing required to verify that the picture on the badge had not been altered. Another area that required considerable work was the high density printed data storage required to get sufficient data on the badge for verification of the picture. The image processing process was successfully tested, and recommendations are included to refine the badge system to ensure high reliability. A two dimensional data array suitable for printing the required data on the badge was proposed, but testing of the readability of the array had to be abandoned due to reallocation of the budgeted funds by the LDRD office

Estimates of relative areas for the disposal in bedded salt of LWR wastes from alternative fuel cycles

International Nuclear Information System (INIS)

Lincoln, R.C.; Larson, D.W.; Sisson, C.E.

1978-01-01

The relative mine-level areas (land use requirements) which would be required for the disposal of light-water reactor (LWR) radioactive wastes in a hypothetical bedded-salt formation have been estimated. Five waste types from alternative fuel cycles have been considered. The relative thermal response of each of five different site conditions to each waste type has been determined. The fuel cycles considered are the once-through (no recycle), the uranium-only recycle, and the uranium and plutonium recycle. The waste types which were considered include (1) unreprocessed spent reactor fuel, (2) solidified waste derived from reprocessing uranium oxide fuel, (3) plutonium recovered from reprocessing spent reactor fuel and doped with 1.5% of the accompanying waste from reprocessing uranium oxide fuel, (4) waste derived from reprocessing mixed uranium/plutonium oxide fuel in the third recycle, and (5) unreprocessed spent fuel after three recycles of mixed uranium/plutonium oxide fuels. The relative waste-disposal areas were determined from a calculated value of maximum thermal energy (MTE) content of the geologic formations. Results are presented for each geologic site condition in terms of area ratios. Disposal area requirements for each waste type are expressed as ratios relative to the smallest area requirement (for waste type No. 2 above). For the reference geologic site condition, the estimated mine-level disposal area ratios are 4.9 for waste type No. 1, 4.3 for No. 3, 2.6 for No. 4, and 11 for No. 5

Health physics documentation

International Nuclear Information System (INIS)

Stablein, G.

1980-01-01

When dealing with radioactive material the health physicist receives innumerable papers and documents within the fields of researching, prosecuting, organizing and justifying radiation protection. Some of these papers are requested by the health physicist and some are required by law. The scope, quantity and deposit periods of the health physics documentation at the Karlsruhe Nuclear Research Center are presented and rationalizing methods discussed. The aim of this documentation should be the application of physics to accident prevention, i.e. documentation should protect those concerned and not the health physicist. (H.K.)

ELECTRICAL SUPPORT SYSTEM DESCRIPTION DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

S. Roy

2004-06-24

The purpose of this revision of the System Design Description (SDD) is to establish requirements that drive the design of the electrical support system and their bases to allow the design effort to proceed to License Application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience/users are design engineers. This type of SDD both ''leads'' and ''trails'' the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. Functional and operational requirements applicable to electrical support systems are obtained from the ''Project Functional and Operational Requirements'' (F&OR) (Siddoway 2003). Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (Doraswamy 2004), and fire hazards analyses. The above-mentioned low-level documents address ''Project Requirements Document'' (PRD) (Canon and Leitner 2003) requirements. This SDD contains several appendices that include supporting information. Appendix B lists key system charts, diagrams, drawings, and lists, and Appendix C includes a list of system procedures.

Study on material attractiveness aspect of spent nuclear fuel of LWR and FBR cycles based on isotopic plutonium production

International Nuclear Information System (INIS)

Permana, Sidik; Suzuki, Mitsutoshi; Saito, Masaki; Novitrian,; Waris, Abdul; Suud, Zaki

2013-01-01

Highlights: • The paper analyzes the plutonium production of recycling nuclear fuel option. • To evaluate material attractiveness based on intrinsic feature of material barrier. • Evaluation based on isotopic plutonium composition of spent fuel LWR and FBR. • Even mass number of plutonium gives a significant contribution to material barrier, in particular Pu-238 and Pu-240. • Doping MA in FBR blanket is effective to increase material barrier from weapon grade plutonium to more than MOX fuel grade. - Abstract: Recycling minor actinide (MA) as well as used uranium and plutonium can be considered to reduce nuclear waste production as well as to increase the intrinsic aspect of nuclear nonproliferation as doping material. Plutonium production as a significant aspect of recycling nuclear fuel option, gives some advantages and challenges, such as fissile material utilization of plutonium as well as production of some even mass number plutonium. The study intends to evaluate the material attractiveness based on the intrinsic feature of material barrier such as plutonium composition, decay heat and spontaneous fission neutron components from spent fuel (SF) light water reactor (LWR) and fast breeder reactor (FBR) cycles. A significant contribution has been shown by decay heat (DH) and spontaneous fission neutron (SFN) of even mass number of plutonium isotopes to the total DH and SFN of plutonium element, in particular from isotopic plutonium Pu-238 and Pu-240 contributions. Longer decay cooling time and higher burnup are effective to increase the material barrier (DH and SFN) level from reactor grade plutonium level to MOX grade plutonium level. Material barrier of plutonium element from spent fuel (SF) FBR in the core regions has similarity to the material barrier profile of SF LWR which can be categorized as MOX fuel grade plutonium. Plutonium compositions, DH and SFN components are categorized as weapon grade plutonium level for FBR blanket regions with no

CCGT + LWR = the power plant of the future?

International Nuclear Information System (INIS)

Tsiklauri, G.

1997-01-01

The thermal efficiency of LWR type reactors can be increased making use of the Tsikl-Durst cycle, where the gas turbine is combined with the nuclear reactor using a steam mixer. The principle of this combined cycle is outlined. It is envisaged that the overall thermal efficiency of the power plant can be increased to 41 - 44%. The total output would be two to three times higher. With advanced light-water reactors (ABWR, AP-600) and advanced gas turbines in combination with the one-way steam generator as developed by Solar Turbines Inc., producing steam at 650 degC to 750 degC, it is feasible to attain a total thermal efficiency of 55%. The combination of two kinds of fuel (nuclear fuel and natural gas) improves operating flexibility of the cycle in various regimes so as to respond to natural gas prices and electricity demands. The gas turbine adds to the nuclear power plant an independent source of power, so that standby dieselgenerators are no more necessary. (P.A.). 1 tab., 2 figs

Adapting LWR to future needs: SECURE-P (PIUS)

International Nuclear Information System (INIS)

Hannerz, K.

1984-01-01

Advanced nuclear technology based on breeder reactors and fuel reprocessing may eventually be applied on a large scale, although the timing for this appears uncertain. However, in many parts of the world societal conditions and technological infrastructure mandate the use of a less complicated technology if the benefits of clean, safe nuclear power are to be available. Such a technology must be based on thermal reactors. Lack of fuel resources for their operation through most of the next century is unlikely to be a serious limitation. A natural contender would be the light water reactor, but today's designs lack many of the desired characteristics. However, introduction of certain new design features can eliminate the shortcomings and make the LWR the prime longterm candidate for a simple, technologically unsophisticated generation of nuclear power. Availability of such an option will also be a major asset for utilities in the large industrial countries before the advent of the era of advanced 'second generation' nuclear power. The costs of demonstrating the new design features are miniscule in relation to the benefits that should accrue. (author)

Democratic People's Republic of Korea LWR project status

International Nuclear Information System (INIS)

Mulligan, J.B.

1996-01-01

In October 1994, at Geneva, the United States and the Democratic People's Republic of Korea (DPRK) signed an Agreed Framework as a first step toward resolving international concerns about nuclear activities in the DPRK. This Agreement, when implemented, will ultimately lead to the complete dismantlement of those aspects of the DPRK's nuclear program, including reprocessing-related facilities, that have undermined the viability of the international nuclear non-proliferation regime and the stability of the Asia-Pacific region. The essence of the Agreement is that the DPRK will take near-term action to cease the activities of concern and permit some International Atomic Energy Agency (IAEA) verification inspection. In the future, it will dismantle its production reactors and accept full-scope IAWA safeguards. In return, the United Stated agreed to lead an international effort to supply the DPRK with light-water reactors which are less of proliferation concern than are graphite-moderated production reactors. Until the first LWR is in operation the DPRK will receive shipments of heavy oil to replace the energy lost by shutting down the production reactors

Characterization and chemistry of fission products released from LWR fuel under accident conditions

International Nuclear Information System (INIS)

Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

1984-01-01

Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

Prioritization of tasks in the draft LWR safety technology program plan. Final report

International Nuclear Information System (INIS)

Lim, E.Y.; Miller, W.J.; Parkinson, W.J.; Ritzman, R.L.; vonHerrmann, J.L.; Wood, P.J.

1980-05-01

The purpose of this report is to describe both the approach taken and the results produced in the SAI effort to prioritize the tasks in the Sandia draft LWR Safety Technology Program Plan. This work used the description of important safety issues developed in the Reactor Safety Study (2) to quantify the effect of safety improvements resulting from a research and development program on the risk from nuclear power plants. Costs of implementation of these safety improvements were also estimated to allow a presentation of the final results in a value (i.e., risk reduction) vs. impact (i.e., implementation costs) matrix

Uncertainties in criticality analysis which affect the storage and transportation of LWR fuel

International Nuclear Information System (INIS)

Napolitani, D.G.

1989-01-01

Satisfying the design criteria for subcriticality with uncertainties affects: the capacity of LWR storage arrays, maximum allowable enrichment, minimum allowable burnup and economics of various storage options. There are uncertainties due to: calculational method, data libraries, geometric limitations, modelling bias, the number and quality of benchmarks performed and mechanical uncertainties in the array. Yankee Atomic Electric Co. (YAEC) has developed and benchmarked methods to handle: high density storage rack designs, pin consolidation, low density moderation and burnup credit. The uncertainties associated with such criticality analysis are quantified on the basis of clean criticals, power reactor criticals and intercomparison of independent analysis methods

Fabrication of Multi-Layerd SiC Composite Tube for LWR Applications

Energy Technology Data Exchange (ETDEWEB)

Kim, Daejong; Jung, Choonghwan; Kim, Weonju; Park, Jiyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Jongmin [Chungnam National Univ., Daejeon (Korea, Republic of)

2013-05-15

In this study, the chemical vapor deposition (CVD) and chemical vapor infiltration (CVI) methods were employed for the fabrication of the composite tubes. SiC ceramics and SiC-based composites have recently been studied for LWR fuel cladding applications because of good mechanical/physical properties, neutron irradiation resistance and excellent compatibility with coolant under severe accident. A multi-layered SiC composite tube as the nuclear fuel cladding is composed of the monolith SiC inner layer, SiC/SiC composite intermediate layer, and monolith SiC outer layer. Since all constituents should be highly pure, stoichiometric to achieve the good properties, it has been considered that the chemical process is a well-suited technique for the fabrication of the SiC phases.

Prediction of fission product and aerosol behaviour during a postulated severe accident in a LWR

International Nuclear Information System (INIS)

Guentay, S.; Aeby, F.; Raguin, M.; Passalacqua, R.

1990-02-01

Lack of appropriate energy removal causes fuel elements in a reactor core to overheat and may eventually cause core to degrade. Fission products will be emitted from a degraded reactor core. Aerosols are generated when the vapours of various fuel and structural materials reach a cold environment and nucleate. In addition to the fission products release and aerosol generation taking place in the reactor vessel, some more fission products release and aerosol generation will occur when the molten core debris leaves the pressure vessel bottom head and comes in contact with the pedestal concrete floor. Fission products, if they are released to environment from the containment boundary, exert a great danger to public health. A source term is defined as the quantity, timing, and characteristics of the release of radionuclide material to the environment following a postulated severe accident. At PSI a considerable effort hase been spent in investigating and establishing a source term assessment methodology in order to predict the source term for a given Light Water Reactor (LWR) accident scenario. This report introduces the computer programs and the methods associated with the release of the fission products, generation of the aerosols and behaviour of the aerosols in LWR compartments used for a source term assessment analysis at PSI. (author) 4 figs., 5 tabs., 28 refs

Information on the evolution of severe LWR fuel element damage obtained in the CORA program

International Nuclear Information System (INIS)

Schanz, G.; Hagen, S.; Hofmann, P.; Sepold, L.; Schumacher, G.

1992-01-01

In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400deg C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4 C), which are leading to extensive low-temperature melt formation around 1200deg C. Interrelations between those basic phenomena, resulting for example in cladding deformation ('flowering') and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ('quenching') are determining the evolution paths of fuel element destruction, which are to be identified. (orig.)

MCNP analysis of the nine-cell LWR gadolinium benchmark

International Nuclear Information System (INIS)

Arkuszewski, J.J.

1988-01-01

The Monte Carlo results for a 9-cell fragment of the light water reactor square lattice with a central gadolinium-loaded pin are presented. The calculations are performed with the code MCNP-3A and the ENDF-B/5 library and compared with the results obtained from the BOXER code system and the JEF-1 library. The objective of this exercise is to study the feasibility of BOXER for the analysis of a Gd-loaded LWR lattice in the broader framework of GAP International Benchmark Analysis. A comparison of results indicates that, apart from unavoidable discrepancies originating from different data evaluations, the BOXER code overestimates the multiplication factor by 1.4 % and underestimates the power release in a Gd cell by 4.66 %. It is hoped that further similar studies with use of the JEF-1 library for both BOXER and MCNP will help to isolate and explain these discrepancies in a cleaner way. (author) 4 refs., 9 figs., 10 tabs

Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

International Nuclear Information System (INIS)

Katayama, Y.B.

1979-04-01

Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

Automatic test equipment for C and I of compact LWR

International Nuclear Information System (INIS)

Mayya, Anuradha; Marathe, P.P.; Madala, Kalyan C.

2014-01-01

The C and I of compact LWR consist of a wide variety of electronic modules. Testing of these modules manually was found to be very cumbersome. To ease the testing of these modules, Automatic Test Equipments (ATE) were developed jointly by BARC and ECIL. This paper describes the design of two ATEs for testing 69 types of modules. A power supply ATE was developed for 43 types of power supply modules of type AC-AC, AC-DC, DC-DC and signal conditioning modules. A VME ATE was developed to test 26 types of VME bus based and other microcontroller based non-bussed modules. These ATEs are used for the automated black box testing of modules by feeding power and control inputs and checking the outputs without operator intervention. This paper describes the important considerations in design and the major design challenges. (author)

Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report

International Nuclear Information System (INIS)

Baldwin, M.N.; Hoovler, G.S.; Eng, R.L.; Welfare, F.G.

1979-07-01

Close-packed storage of LWR fuel assemblies is needed in order to expand the capacity of existing underwater storage pools. This increased capacity is required to accommodate the large volume of spent fuel produced by prolonged onsite storage. To provide benchmark criticality data in support of this effort, 20 critical assemblies were constructed that simulated a variety of close-packed LWR fuel storage configurations. Criticality calculations using the Monte Carlo KENO-IV code were performed to provide an analytical basis for comparison with the experimental data. Each critical configuration is documented in sufficient detail to permit the use of these data in validating calculational methods according to ANSI Standard N16.9-1975

Evaluation of the 252Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

International Nuclear Information System (INIS)

Mihalczo, J.T.

1987-01-01

The 252 Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt LWR fuel submerged in fuel storage pools, fully loaded and as they are being loaded. The motivation for this evaluation was that measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This in turn could lead to more cost-effective cask designs. Evaluation of the method for this application was based on (1) experiments already completed at a critical experiments facility using arrays of PWR fuel pins typical of the size of storage cask configurations, (2) the existence of neutron detectors that can function in shipping cask environments, and (3) the ability to construct ionization chambers containing 252 Cf of adequate intensity for these measurements. These three considerations are discussed

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

International Nuclear Information System (INIS)

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.

Program plan for research and development in support of LWR fuel recycle

International Nuclear Information System (INIS)

1975-01-01

The ERDA program that is being planned to assist industry in the commercialization of the LWR fuel cycle will involve a range of activities, including joint programs with industry, R and D to provide technology, conceptual design of fuel recycle facilities, and environmental and economic assessments. A two-part program to begin in 1976 that is a portion of the overall ERDA plan is described. Responsibility for coordination and management of the tasks described in this document has been assigned to Du Pont as prime contractor to the ERDA Savannah River Operations Office. The first part of the program consists of the conceptual design of complete recycle facilities. The second part of the program, which will proceed concurrently, consists of supporting R and D activities, economic and environmental studies, and other studies to assist in the regulatory process. The R and D program will include both near-term activities in support of the conceptual design effort, and other activities aimed at general improvements in fuel cycle technology. The conceptual design will be used to develop current cost information for a complete reprocessing complex. The design will be based initially on current technology with provision for improvements as confirmatory information and advanced technology become available from the R and D program. The conceptual design and cost estimate will be developed by the Du Pont Atomic Energy Division. The R and D program and supporting studies will be directed at uncertainties in current technology as well as toward development of improved technology. It will include such R and D as might be appropriate for ERDA to undertake in support of joint programs with industry. The Savannah River Laboratory will have responsibility for coordinating the program

Behavior of LWR fuel elements under accident conditions

International Nuclear Information System (INIS)

Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

1977-01-01

In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

Technical support document for proposed 1994 revision of the MEC thermal envelope requirements

Energy Technology Data Exchange (ETDEWEB)

Conner, C.C.; Lucas, R.G.

1994-03-01

This report documents the development of the proposed revision of the Council of American Building Officials` (CABO) 1994 supplement to the 1993 Model Energy Code (MEC) building thermal envelope requirements for maximum component U{sub 0}-value. The 1994 amendments to the 1993 MEC were established in last year`s code change cycle and did not change the envelope requirements. The research underlying the proposed MEC revision was conducted by Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE) Building Energy Standards program. The goal of this research was to develop revised guidelines based on an objective methodology that determines the most cost-effective (least total cost) combination of energy conservation measures (ECMs) (insulation levels and window types) for residential buildings. This least-cost set of ECMs was used as a basis for proposing revised MEC maximum U{sub 0}-values (thermal transmittances). ECMs include window types (for example, double-pane vinyl) and insulation levels (for example, R-19) for ceilings, walls, and floors.

Selection methodology for LWR safety programs and proposals. Volume 2. Methodology application

International Nuclear Information System (INIS)

Ritzman, R.L.; Husseiny, A.A.

1980-08-01

The results of work done to update and apply a methodology for selecting (prioritizing) LWR safety technology R and D programs are described. The methodology is based on multiattribute utility (MAU) theory. Application of the methodology to rank-order a group of specific R and D programs included development of a complete set of attribute utility functions, specification of individual attribute scaling constants, and refinement and use of an interactive computer program (MAUP) to process decision-maker inputs and generate overall (multiattribute) program utility values. The output results from several decision-makers are examined for consistency and conclusions and recommendations regarding general use of the methodology are presented. 3 figures, 18 tables

Molten Salt Reactor Experiment Facility (Building 7503) standards/requirements identification document adherence assessment plan at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1996-02-01

This is the Phase 2 (adherence) assessment plan for the Building 7503 Molten Salt Reactor Experiment (MSRE) Facility standards/requirements identification document (S/RID). This document outlines the activities to be conducted from FY 1996 through FY 1998 to ensure that the standards and requirements identified in the MSRE S/RID are being implemented properly. This plan is required in accordance with the Department of Energy Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 90-2, November 9, 1994, Attachment 1A. This plan addresses the major aspects of the adherence assessment and will be consistent with Energy Systems procedure QA-2. 7 ''Surveillances.''

LWR primary coolant pipe rupture test rig

International Nuclear Information System (INIS)

Yoshitoshi, Shyoji

1978-01-01

The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

Energy Technology Data Exchange (ETDEWEB)

Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

1983-06-01

LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States); Mei, Zhigang [Argonne National Lab. (ANL), Argonne, IL (United States); Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-12-10

Uranium silicide (U₃Si₂) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U₃Si₂, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U₃Si₂ at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuel material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U₃Si₂ in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U₃Si₂ at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U₃Si₂ fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U₃Si₂ fuel as an accident

Assessment of nitrogen as an atmosphere for dry storage of spent LWR fuel

International Nuclear Information System (INIS)

Gilbert, E.R.; Knox, C.A.; White, G.D.

1985-09-01

Interim dry storage of spent light-water reactor (LWR) fuel is being developed as a licensed technology in the United States. Because it is anticipated that license agreements will specify dry storage atmospheres, the behavior of spent LWR fuel in a nitrogen atmosphere during dry storage was investigated. In particular, the thermodynamics of reaction of nitrogen compounds (expected to form in the cover gas during dry storage) and residual impurities (such as moisture and oxygen) with Zircaloy cladding and with spent fuel at sites of cladding breaches were examined. The kinetics of reaction were not considered it was assumed that the 20 to 40 years of interim dry storage would be sufficient for reactions to proceed to completion. The primary thermodynamics reactants were found to be NO 2 , N 2 O, H 2 O 2 , and O 2 . The evaluation revealed that the limited inventories of these reactants produced by the source terms in hermetically sealed dry storage systems would be too low to cause significant spent fuel degradation. Furthermore, the oxidation of spent fuel to degrading O/U ratios is unlikely because the oxidation potential in moist nitrogen limits O/U ratios to values less than UO/sub 2.006/ (the equilibrium stoichiometric form in equilibrium with moist nitrogen). Tests were performed with bare spent UO 2 fuel and nonirradiated UO 2 pellets (with no Zircaloy cladding) in a nitrogen atmosphere containing moisture concentrations greater than encountered under dry storage conditions. These tests were performed for at least 1100 h at temperatures as high as 380 0 C, where oxidation reactions proceed in a matter of minutes. No visible degradation was detected, and weight changes were negligible

Regulatory document R-104, Regulatory objectives, requirements and guidelines for the disposal of radioactive wastes - long-term aspects

International Nuclear Information System (INIS)

1987-01-01

The purpose and scope of this document is to present the regulatory basis for judging the long-term acceptability of radioactive waste disposal options. The basic objectives of radioactive waste disposal are given as are the regulatory requirements to be satisfied. (NEA)

Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

International Nuclear Information System (INIS)

Black, J.E.; King, F.D.

1979-01-01

Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

Tank Monitoring and Document control System (TMACS) As Built Software Design Document

International Nuclear Information System (INIS)

GLASSCOCK, J.A.

2000-01-01

This document describes the software design for the Tank Monitor and Control System (TMACS). This document captures the existing as-built design of TMACS as of November 1999. It will be used as a reference document to the system maintainers who will be maintaining and modifying the TMACS functions as necessary. The heart of the TMACS system is the ''point-processing'' functionality where a sample value is received from the field sensors and the value is analyzed, logged, or alarmed as required. This Software Design Document focuses on the point-processing functions

CASK/MSC/WP PREPARATION SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

S. Drummond

2005-01-01

The purpose of this system description document (SDD) is to establish requirements that drive the design of the Cask/MSC/WP preparation system and their bases to allow the design effort to proceed to license application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design, as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This type of SDD both leads and trails the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. This SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. This SDD addresses the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]) requirements. Additional PRD requirements may be cited, as applicable, to drive the design of specific aspects of the system, with justifications provided in the basis. Functional and operational requirements applicable to this system are obtained from the ''Project Functional and Operational Requirements'' (F and OR) (Curry 2004 [DIRS 170557]) document. Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (BSC 2004 [DIRS 171599]) and the preclosure safety analyses

SNF AGING SYSTEM DESCRIPTION DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

L.L. Swanson

2005-04-06

The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts

SNF AGING SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

L.L. Swanson

2005-01-01

The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F andOR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts, diagrams, drawings, lists and additional supporting information; and Appendix C is a list of

Documentation of medical findings in radiation workers in the GDR to meet the requirements of ICRP publication 26

International Nuclear Information System (INIS)

Wolff, H.R.; Neumeister, K.

1979-01-01

Based on ICRP Publication 26, the future organization of the medical surveillance system for radiation workers in the GDR is considered in this paper. These radiation workers will also in future be medically supervised by means of pre-employment and routine examinations. It is considered necessary to have as extensive a registration as possible of information on medical examinations, working place analyses and incidents. Such data have to be collected and stored to be compared with other national and international projects (e.g. in the field of occupational health). In addition, they should permit epidemiological studies to be internationally co-ordinated. For this purpose, a documentation system has been prepared in the German Democratic Republic which is based on GDR experiences and makes it possible to specify the requirements of ICRP Publication 26. This system forms a new basis for mass examinations of occupationally exposed persons. Uniform examination methods tailored to meet the task of assuring occupational health in the GDR will be introduced. The documentation cards are meant to be used as clear-text cards suited for automatic reading by optical character recognition. The examination form consists of ten parts and comprises all details from working place situation to medical findings to laboratory results. It is felt that this new documentation system permits registration of all relevant data required for the effective radiation protection of man. On the basis of this documentation of findings, participation is scheduled in the respective international IAEA programmes and the studies proposed by the ICRP for problems of radiation-induced carcinogenesis and radiogenetics

Development of PIE techniques for irradiated LWR pressure vessel steels

International Nuclear Information System (INIS)

Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

1999-01-01

For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

Characterization of LWR fuel rod irradiations with power transients in the BR2 reflector

International Nuclear Information System (INIS)

Ponsard, B.; Bodart, S.; Meer, K. van der; Raedt, C. de

1996-01-01

Fuel rod irradiations in reflector positions of the materials testing reactor BR2 are becoming increasingly important. A typical example is that of irradiation devices containing single LWR fuel rods, to be tested in the framework of a new international fuel investigation and development programme. Some of the irradiations will comprise power transients with central fuel melting (at 2800 deg. C), the power increase being obtained by decreasing the pressure in a He-3 neutron absorbing screen and/or by varying the BR2 reactor operating power. A total power variation by a factor of at least 2.5 in the fuel rod irradiated could thus be achieved. In some of the rods, central temperature measurements (up to 2000 deg. C) will be carried out. Both fresh and pre-irradiated fuel rods are concerned in the programme. For these irradiations, the accurate knowledge of the neutron-induced fission heating and of the gamma heating is required, as one of the purposes of the programme consists in establishing the correlation among the thermal conductivity, the burn-up and the irradiation temperature. Calibration work among various measuring methods and between measurements and one- and two-dimensional calculations is being pursued. (author). 10 refs, 15 figs, 3 tabs

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 4

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

Radiation protection of personnel and the public is accomplished by establishing a well defined Radiation Protection Organization to ensure that appropriate controls on radioactive materials and radiation sources are implemented and documented. This Requirements Identification Document (RID) applies to the activities, personnel, structures, systems, components, and programs involved in executing the mission of the Tank Farms. The physical boundaries within which the requirements of this RID apply are the Single Shell Tank Farms, Double Shell Tank Farms, 242-A Evaporator-Crystallizer, 242-S, T Evaporators, Liquid Effluent Retention Facility (LERF), Purgewater Storage Facility (PWSF), and all interconnecting piping, valves, instrumentation, and controls. Also included is all piping, valves, instrumentation, and controls up to and including the most remote valve under Tank Farms control at any other Hanford Facility having an interconnection with Tank Farms. The boundary of the structures, systems, components, and programs to which this RID applies, is defined by those that are dedicated to and/or under the control of the Tank Farms Operations Department and are specifically implemented at the Tank Farms.

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 4

International Nuclear Information System (INIS)

1994-04-01

Radiation protection of personnel and the public is accomplished by establishing a well defined Radiation Protection Organization to ensure that appropriate controls on radioactive materials and radiation sources are implemented and documented. This Requirements Identification Document (RID) applies to the activities, personnel, structures, systems, components, and programs involved in executing the mission of the Tank Farms. The physical boundaries within which the requirements of this RID apply are the Single Shell Tank Farms, Double Shell Tank Farms, 242-A Evaporator-Crystallizer, 242-S, T Evaporators, Liquid Effluent Retention Facility (LERF), Purgewater Storage Facility (PWSF), and all interconnecting piping, valves, instrumentation, and controls. Also included is all piping, valves, instrumentation, and controls up to and including the most remote valve under Tank Farms control at any other Hanford Facility having an interconnection with Tank Farms. The boundary of the structures, systems, components, and programs to which this RID applies, is defined by those that are dedicated to and/or under the control of the Tank Farms Operations Department and are specifically implemented at the Tank Farms

Impact of implicit effects on uncertainties and sensitivities of the Doppler coefficient of a LWR pin cell

Science.gov (United States)

Hursin, Mathieu; Leray, Olivier; Perret, Gregory; Pautz, Andreas; Bostelmann, Friederike; Aures, Alexander; Zwermann, Winfried

2017-09-01

In the present work, PSI and GRS sensitivity analysis (SA) and uncertainty quantification (UQ) methods, SHARK-X and XSUSA respectively, are compared for reactivity coefficient calculation; for reference the results of the TSUNAMI and SAMPLER modules of the SCALE code package are also provided. The main objective of paper is to assess the impact of the implicit effect, e.g., considering the effect of cross section perturbation on the self-shielding calculation, on the Doppler coefficient SA and UQ. Analyses are done for a Light Water Reactor (LWR) pin cell based on Phase I of the UAM LWR benchmark. The negligence of implicit effects in XSUSA and TSUNAMI leads to deviations of a few percent between the sensitivity profiles compared to SAMPLER and TSUNAMI (incl. implicit effects) except for 238U elastic scattering. The implicit effect is much larger for the SHARK-X calculations because of its coarser energy group structure between 10 eV and 10 keV compared to the applied SCALE libraries. It is concluded that the influence of the implicit effect strongly depends on the energy mesh of the nuclear data library of the neutron transport solver involved in the UQ calculations and may be magnified by the response considered.

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 1

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

The purpose of this Requirements Identification Document (RID) section is to identify, in one location, all of the facility specific requirements and good industry practices which are necessary or important to establish an effective Issues Management Program for the Tank Farm Facility. The Management Systems Functional Area includes the site management commitment to environmental safety and health (ES&H) policies and controls, to compliance management, to development and management of policy and procedures, to occurrence reporting and corrective actions, resource and issue management, and to the self-assessment process.

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 1

International Nuclear Information System (INIS)

1994-04-01

The purpose of this Requirements Identification Document (RID) section is to identify, in one location, all of the facility specific requirements and good industry practices which are necessary or important to establish an effective Issues Management Program for the Tank Farm Facility. The Management Systems Functional Area includes the site management commitment to environmental safety and health (ES ampersand H) policies and controls, to compliance management, to development and management of policy and procedures, to occurrence reporting and corrective actions, resource and issue management, and to the self-assessment process

TECHNICAL BASIS FOR VENTILATION REQUIREMENTS IN TANK FARMS OPERATING SPECIFICATIONS DOCUMENTS

Energy Technology Data Exchange (ETDEWEB)

BERGLIN, E J

2003-06-23

This report provides the technical basis for high efficiency particulate air filter (HEPA) for Hanford tank farm ventilation systems (sometimes known as heating, ventilation and air conditioning [HVAC]) to support limits defined in Process Engineering Operating Specification Documents (OSDs). This technical basis included a review of older technical basis and provides clarifications, as necessary, to technical basis limit revisions or justification. This document provides an updated technical basis for tank farm ventilation systems related to Operation Specification Documents (OSDs) for double-shell tanks (DSTs), single-shell tanks (SSTs), double-contained receiver tanks (DCRTs), catch tanks, and various other miscellaneous facilities.

CASK/MSC/WP PREPARATION SYSTEM DESCRIPTION DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

S. Drummond

2005-04-12

The purpose of this system description document (SDD) is to establish requirements that drive the design of the Cask/MSC/WP preparation system and their bases to allow the design effort to proceed to license application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design, as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This type of SDD both leads and trails the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. This SDD trails the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. This SDD addresses the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]) requirements. Additional PRD requirements may be cited, as applicable, to drive the design of specific aspects of the system, with justifications provided in the basis. Functional and operational requirements applicable to this system are obtained from the ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]) document. Other requirements to support the design process have been taken from higher-level requirements documents such as the ''Project Design Criteria Document'' (PDC) (BSC 2004 [DIRS 171599]) and the preclosure safety analyses.

Perspective on US NRC Policy Issues Concerning Use of Risk Insights for Non-LWR

International Nuclear Information System (INIS)

Ha, Jun Su; Kim, In Goo; Huh, Chang Wook; Kim, Kyun Tae

2011-01-01

Since the PRA Implementation plan of US NRC (1994), PRA has been applied to all NPPs in USA and risk insights have been used for the regulation as a complement of the deterministic approaches. RIRIP (Risk-Informed Regulation Implementation Plan, 2000) and RPP (Risk-Informed and Performance-Based Plan, 2007) were announced by US NRC thereafter, which recommended enhanced use of risk insights. In the meantime, there have been lots of policy issues concerning use of risk insights for licensing Non-LWR designs, which will be discussed in this paper to understand the stream of perspectives on US NRC's approach

Risk management: integration of social and technical risk variables into safety assessments of LWR'S

International Nuclear Information System (INIS)

Turnage, J.J.; Husseiny, A.A.

1980-01-01

A risk management methodology is developed here to formalize the acceptability levels of commercial LWR power plants via the estimation of risk levels acceptable to the public and the integration of such estimates into risk-benefit analysis. Utility theory is used for developing preference models based on value trade-offs among multiple objectives and uncertainties about the impact of alternatives. The method involves reducing the various variables affecting safety acceptability decisions to a single function that provides a metric for acceptability levels. The function accomondates for technical criteria related to design and licensing decisions, as well as public reactions to certain choices

Advantages of retrofitting high velocity separators to LWR turbines; experience in VVR NPP Loviisa

International Nuclear Information System (INIS)

Dueymes, E.; Peyrelongue, J.P.

1992-01-01

Erosion-corrosion by wet steam is a concern for VVER operators and also, in numerous LWR power plants of western technology. The backfitting of moisture separators at the HP Turbine outlets is a way to avoid maintenance costs, repairs, replacement of pipes or equipments. Installation of HVS at LOVIISA confirms that this device, whose installation work is reduced to a minimum, is able to remove quite all the water from the steam just a few meters downstream the HP cylinder. A long term operation can be expected for carbon steel equipments, even those previously damaged by erosion-corrosion. (authors). 6 figs., 2 tabs

NASA software documentation standard software engineering program

Science.gov (United States)

1991-01-01

The NASA Software Documentation Standard (hereinafter referred to as Standard) can be applied to the documentation of all NASA software. This Standard is limited to documentation format and content requirements. It does not mandate specific management, engineering, or assurance standards or techniques. This Standard defines the format and content of documentation for software acquisition, development, and sustaining engineering. Format requirements address where information shall be recorded and content requirements address what information shall be recorded. This Standard provides a framework to allow consistency of documentation across NASA and visibility into the completeness of project documentation. This basic framework consists of four major sections (or volumes). The Management Plan contains all planning and business aspects of a software project, including engineering and assurance planning. The Product Specification contains all technical engineering information, including software requirements and design. The Assurance and Test Procedures contains all technical assurance information, including Test, Quality Assurance (QA), and Verification and Validation (V&V). The Management, Engineering, and Assurance Reports is the library and/or listing of all project reports.

Tank Monitoring and Document control System (TMACS) As Built Software Design Document

Energy Technology Data Exchange (ETDEWEB)

GLASSCOCK, J.A.

2000-01-27

This document describes the software design for the Tank Monitor and Control System (TMACS). This document captures the existing as-built design of TMACS as of November 1999. It will be used as a reference document to the system maintainers who will be maintaining and modifying the TMACS functions as necessary. The heart of the TMACS system is the ''point-processing'' functionality where a sample value is received from the field sensors and the value is analyzed, logged, or alarmed as required. This Software Design Document focuses on the point-processing functions.

Stress corrosion cracking of age-hardenable nickel-base alloys in LWR-conditions

International Nuclear Information System (INIS)

Kekkonen, T.; Haenninen, H.

1985-01-01

At present it seems that the microstructure most resistant to stress corrosion cracking (SCC) in high temperature water is obtained by a solution annealing treatment at a relatively high temperature (appr. 1100 deg C) followed by water quenching and a single aging treatment (appr. 700 deg C/20 h). This should produce a microstructure with a high M 23 Cc 6 :MC ratio, semi-continous coherent M 23 C 6 precipitation, and an evenly distributed gamma prime in the matrix. However, since the actual mechanism of SCC in age-hardenable Ni-base alloys is unclear, the microstructural features resulting in the good resistance to SCC cannot be specified. Furthermore, the possible microstructural changes caused by prolonged use in LWR-conditions are unknown

Proposal of a nuclear cycle research and development plan in Tokai works. The roadmap from LWR cycle to FBR cycle

International Nuclear Information System (INIS)

Nakamura, Hirofumi; Abe, Tomoyuki; Kashimura, Takuo; Nagai, Toshihisa; Maeda, Seichiro; Yamaguchi, Toshiya; Kuroki, Ryoichiro

2003-07-01

The Generation-II Project Task Force Team has investigated a research and development plan of a future nuclear fuel cycle in Tokai works for about three months from December 19, 2002. First we have discussed about the present condition of Japanese nuclear fuel cycle and have recognized it as the following. The relation of the technology between the LWR-cycle and the FBR-cycle is not clear. MOX Fuel Use in Light Water Reactors is important to establish technology of the FBR fuel cycle. Radioactive waste disposal issue is urgent. Next we have proposed the three basic policies on R and D plan of nuclear fuel cycle in consideration of the F.S. on FBR-cycle. Establishment and advancement of 'the tough nuclear fuel cycle'. Early establishment of the FBR cycle technology to be able to supply energy stably for long-term. Establishment of the radioactive waste treatment and disposal technology, and optimization of nuclear fuel cycle technology from the viewpoint of radioactive waste. And we have proposed the Japanese technical holder system to integrate all LWR and FBR cycle technology. (author)

LWR reactivity/isotopics code for pedagogical and scoping applications

International Nuclear Information System (INIS)

AbuZaied, G.; Driscoll, M.J.

1986-01-01

A program designated BRICC (Burnup Reactivity and Isotopic Composition Computation), has been programmed for use on microcomputers to permit rapid parametric studies of the neutronics of light water reactor (LWR) assemblies. It is currently employed as a teaching tool in a graduate-level subject on nuclear fuel management, and has proven to be of sufficient accuracy to permit its use as a submodule in a more comprehensive program used to evaluate various mechanical spectral shift concepts for pressurized water reactor control. It should also prove useful in teaching reactor physics as it will fill an important gap between hand calculations of inadequate accuracy and state-of-the-art multigroup programs of daunting complexity. The BRICC program combines a minimum adequate set of old-fashioned phenomenological submodels that describe key physics attributed in an integral fashion, thereby providing the student or researcher with convenient mental pictures to serve as the basis for deductive reasoning. The program is short, written in a simplistic version of the Basic language, with many interspersed Remark statements, and is therefore easy to tinker with for various constructive purposes

Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

International Nuclear Information System (INIS)

Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

1988-12-01

This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

50 CFR 300.185 - Documentation, reporting and recordkeeping requirements for consignment documents and re-export...

Science.gov (United States)

2010-10-01

... of presenting entry documentation for clearance by customs authorities (e.g., CBP Forms 7533 or 3461... apply to all imports of fish or fish products regulated under this subpart, into the Customs territory... landed overseas (HTS heading 9815). For insular possessions with customs territories separate from the...

Critical corrosion issues and mitigation strategies impacting the operability of LWR's

International Nuclear Information System (INIS)

Jones, R.L.

1996-01-01

Recent corrosion experience in US light water reactor nuclear power plants is reviewed with emphasis on mitigation strategies to control the cost of corrosion to LWR operators. Many components have suffered corrosion problems resulting in industry costs of billions of dollars. The most costly issues have been stress corrosion cracking of stainless steel coolant piping in boiling water reactors and corrosion damage to steam generator tubes in pressurized water reactors. Through industry wide R and D programs these problems are now understood and mitigation strategies have been developed to address the issues in a cost effective manner. Other significant corrosion problems for both reactor types are briefly reviewed. Tremendous progress has been made in controlling corrosion, however, minimizing its impact on plant operations will present a continuing challenge throughout the remaining service lives of these power plants

Criticality impacts on LWR fuel storage efficiency

International Nuclear Information System (INIS)

Napolitano, D.

1992-01-01

This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

Document management in engineering construction

International Nuclear Information System (INIS)

Liao Bing

2008-01-01

Document management is one important part of systematic quality management, which is one of the key factors to ensure the construction quality. In the engineering construction, quality management and document management shall interwork all the time, to ensure the construction quality. Quality management ensures that the document is correctly generated and adopted, and thus the completeness, accuracy and systematicness of the document satisfy the filing requirements. Document management ensures that the document is correctly transferred during the construction, and various testimonies such as files and records are kept for the engineering construction and its quality management. This paper addresses the document management in the engineering construction based on the interwork of the quality management and document management. (author)

A study on the behavior of defected LWR spent fuel

International Nuclear Information System (INIS)

You, Gil Sung; Kim, Eun Ka; Kim, Keon Sik; Suh, Hang Suck; Kim, Seung Jung; Ro, Seung Gy; Park, Chong Mook; Ji, Pyung Gook

1992-03-01

To investigate the storage behavior of the defective LWR spent fuel rods, the characteristic changes of fuel and cladding are to be measured and analyzed. In addition, the oxidation study in air on non-irradiated and irradiated U0 2 was performed. No changes were observed in the tested fuel rods after 30 month storage. The Cs-134, 137 released rapidly during the initial 3 months of storage, but remained in constant value after 3 month storage and the release was almost ceased after 30 month storage. The weight gain of non-irradiated U0 2 samples showed a trend of S type curves and the activation energies were 11OKJ/mol above 350 deg C. and 143KJ/mol below 350 deg C. But irradiated U0 2 showed a rapid increase at initial stage of oxidation and a decrease at later stage when compared with the results of non-irradiated U0 2 . (Author)

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 5

Energy Technology Data Exchange (ETDEWEB)

1994-04-01

The Fire Protection functional area for the Hanford Site Tank Farm facilities and support structures is based on the application of relevant DOE orders, regulations, and industry codes and standards. The fire protection program defined in this document may be divided into three areas: (1) organizational, (2) administrative programmatic features, and (3) technical features. The information presented in each section is in the form of program elements and orders, regulations, industry codes, and standards that serve as the attributes of a fire protection program for the Tank Farm facilities. Upon completion this document will be utilized as the basis to evaluate compliance of the fire protection program being implemented for the Tank Farm facilities with the requirements of DOE orders and industry codes and standards.

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 5

International Nuclear Information System (INIS)

1994-04-01

The Fire Protection functional area for the Hanford Site Tank Farm facilities and support structures is based on the application of relevant DOE orders, regulations, and industry codes and standards. The fire protection program defined in this document may be divided into three areas: (1) organizational, (2) administrative programmatic features, and (3) technical features. The information presented in each section is in the form of program elements and orders, regulations, industry codes, and standards that serve as the attributes of a fire protection program for the Tank Farm facilities. Upon completion this document will be utilized as the basis to evaluate compliance of the fire protection program being implemented for the Tank Farm facilities with the requirements of DOE orders and industry codes and standards

The European Utility Requirements (EUR). Status and near term activities

International Nuclear Information System (INIS)

Berbey, Pierre; Hedin, Francois

2010-01-01

In 1991 5 major European Utilities participating in the US ALWR program decided to develop together a common specification that would contribute to keep the nuclear option open. The European Utility Requirements (EUR) are addressed to the designers and suppliers of LWR plants in order to allow the development of standards designs that can be build and licensed in several European countries with only minor variations. The EUR organization has kept enlarging; today 16 utilities are members of the EUR organization. Seven compliance analyses dedicated respectively to the BWR90, EPR, EPP, ABWR, SWR1000, AP1000 and to the AES92 projects have been already published. The revised version of the EPR subset of the EUR volume 3 was finalized in mid 2009. New LWR projects of potential interest for the EUR utilities are being contemplated. For instance a preliminary assessment of compliance of MHI's APWR project has been worked out in the first months of 2008. Recently EUR organization has decided to launch coordinated actions with other industry groups and other stakeholders. In particular EUR and ENISS organizations have decided to join their efforts in their relations with the IAEA and WENRA organizations with respect to the LWR Gen3 designs. In addition EUR and CORDEL (Cooperation in Reactor Design Evaluation and Licensing), which is a WNA (World Nuclear Association) working group decided also to coordinate their efforts for the industry benefit, in relation with the MDEP (Multinational Design Evaluation Program) initiative of safety nuclear regulators. Contacts have been also initiated with ENEN and the WNU in order to develop new courses for young professionals. (orig.)

Review of provisions on corrosion fatigue and stress corrosion in WWER and Western LWR Codes and Standards

International Nuclear Information System (INIS)

Buckthorpe, D.; Filatov, V.; Tashkinov, A.; Evropin, S.V.; Matocha, K.; Guinovart, J.

2003-01-01

Results are presented from a collaborative project performed on behalf of the European Commission, Working Group Codes and Standards. The work covered the contents of current codes and standards, plant experience and R and D results. Current fatigue design rules use S-N curves based on tests in air. Although WWER and LWR design curves are often similar they are derived, presented and used in different ways and it is neither convenient nor appropriate to harmonise them. Similarly the fatigue crack growth laws used in the various design and in-service inspection rules differ significantly with respect to both growth rates in air and the effects of water reactor environments. Harmonised approaches to the effects of WWER and LWR environments are possible based on results from R and D programmes carried out over the last decade. For carbon and low alloy steels a consistent approach to both crack initiation and growth can be formulated based on the superposition of environmentally assisted cracking effects on the fatigue crack development. The approach indicates that effects of the water environment are minimal given appropriate control of the oxygen content of the water and/or the sulphur content of the steel. For austenitic stainless steels a different mechanisms may apply and a harmonised approach is possible at present only for S-N curves. Although substantial progress has been made with respect to corrosion fatigue, more data and a clearer understanding are required in order to write code provisions particularly in the area of high cycle fatigue. Reactor operation experience shows stress corrosion cracking of austenitic steels is the most common cause of failure. These failures are associated with high residual stresses combined with high levels of dissolved oxygen or the presence of contaminants. For primary circuit internals there is a potential threat to integrity from irradiated assisted stress corrosion cracking. Design and in-service inspection rules do not at

Monitored Geologic Repository Project Description Document

International Nuclear Information System (INIS)

Curry, P.

2000-01-01

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 and 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 and 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)

Repository-based software engineering program: Concept document

Science.gov (United States)

1992-01-01

This document provides the context for Repository-Based Software Engineering's (RBSE's) evolving functional and operational product requirements, and it is the parent document for development of detailed technical and management plans. When furnished, requirements documents will serve as the governing RBSE product specification. The RBSE Program Management Plan will define resources, schedules, and technical and organizational approaches to fulfilling the goals and objectives of this concept. The purpose of this document is to provide a concise overview of RBSE, describe the rationale for the RBSE Program, and define a clear, common vision for RBSE team members and customers. The document also provides the foundation for developing RBSE user and system requirements and a corresponding Program Management Plan. The concept is used to express the program mission to RBSE users and managers and to provide an exhibit for community review.

FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2005-03-03

This document describes the qualitative evaluation of frequency and consequences for DST and SST representative flammable gas accidents and associated hazardous conditions without controls. The evaluation indicated that safety-significant structures, systems and components (SSCs) and/or technical safety requirements (TSRs) were required to prevent or mitigate flammable gas accidents. Discussion on the resulting control decisions is included. This technical basis document was developed to support WP-13033, Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process for the flammable gas representative accidents and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the event frequency and consequence.

A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

Directory of Open Access Journals (Sweden)

Jorge Pérez Mañes

2014-01-01

Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

New Developments in Actinides Burning with Symbiotic LWR-HTR-GCFR Fuel Cycles

International Nuclear Information System (INIS)

Bomboni, Eleonora

2008-01-01

The long-term radiotoxicity of the final waste is currently the main drawback of nuclear power production. Particularly, isotopes of Neptunium and Plutonium along with some long-lived fission products are dangerous for more than 100000 years. 96% of spent Light Water Reactor (LWR) fuel consists of actinides, hence it is able to produce a lot of energy by fission if recycled. Goals of Generation IV Initiative are reduction of long-term radiotoxicity of waste to be stored in geological repositories, a better exploitation of nuclear fuel resources and proliferation resistance. Actually, all these issues are intrinsically connected with each other. It is quite clear that these goals can be achieved only by combining different concepts of Gen. IV nuclear cores in a 'symbiotic' way. Light-Water Reactor - (Very) High Temperature Reactor ((V)HTR) - Fast Reactor (FR) symbiotic cycles have good capabilities from the viewpoints mentioned above. Particularly, HTR fuelled by Plutonium oxide is able to reach an ultra-high burn-up and to burn Neptunium and Plutonium effectively. In contrast, not negligible amounts of Americium and Curium build up in this core, although the total mass of Heavy Metals (HM) is reduced. Americium and Curium are characterised by an high radiological hazard as well. Nevertheless, at least Plutonium from HTR (rich in non-fissile nuclides) and, if appropriate, Americium can be used as fuel for Fast Reactors. If necessary, dedicated assemblies for Minor Actinides (MA) burning can be inserted in Fast Reactors cores. This presentation focuses on combining HTR and Gas Cooled Fast Reactor (GCFR) concepts, fuelled by spent LWR fuel and depleted uranium if need be, to obtain a net reduction of total mass and radiotoxicity of final waste. The intrinsic proliferation resistance of this cycle is highlighted as well. Additionally, some hints about possible Curium management strategies are supplied. Besides, a preliminary assessment of different chemical forms of

Documentation of spectrom-32

International Nuclear Information System (INIS)

Callahan, G.D.; Fossum, A.F.; Svalstad, D.K.

1989-01-01

SPECTROM-32 is a finite element program for analyzing two-dimensional and axisymmetric inelastic thermomechanical problems related to the geological disposal of nuclear waste. The code is part of the SPECTROM series of special-purpose computer programs that are being developed by RE/SPEC Inc. to address many unique rock mechanics problems encountered in analyzing radioactive wastes stored in geologic formations. This document presents the theoretical basis for the mathematical models, the finite element formulation and solution procedure of the program, a description of the input data for the program, verification problems, and details about program support and continuing documentation. The computer code documentation is intended to satisfy the requirements and guidelines outlined in the document entitled Final Technical Position on Documentation of Computer Codes for High-Level Waste Management. The principal component models used in the program involve thermoelastic, thermoviscoelastic, thermoelastic-plastic, and thermoviscoplastic types of material behavior. Special material considerations provide for the incorporation of limited-tension material behavior and consideration of jointed material behavior. Numerous program options provide the capabilities for various boundary conditions, sliding interfaces, excavation, backfill, arbitrary initial stresses, multiple material domains, load incrementation, plotting database storage and access of results, and other features unique to the geologic disposal of radioactive wastes. Numerous verification problems that exercise many of the program options and illustrate the required data input and printed results are included in the documentation

A review on future trends of LWR fuel cycle costs

International Nuclear Information System (INIS)

Tamiya, S.; Otomo, T.; Meguro, T.

1977-01-01

In the cost estimations in the past, the main components of fuel cycle were mining and milling, uranium enrichment and fuel fabrication, and reprocessing charge deemed to be recovered by plutonium credit. Since the oil crisis, every component of fuel cycle cost has gone up in recent years as well as the construction cost of a power station. Recent analysis shows that the costs in the back end of fuel cycle are much higher than those anticipated several years ago, although their contribution to the electricity generating cost by nuclear would be small. The situation of the back end of the fuel cycle has been quite changed in recent years, and there are still many uncertainties in this field, that is, regulatory requirements for reprocessing plant such as safety, safeguards, environmental protection and high level waste management. So, it makes it more difficult to estimate the investment in this sector of fuel cycle, therefore, to estimate the cost of this sector. The institutional problems must be cleared in relation to the ultimate disposal of high level waste, too. Co-location of some parts of fuel cycle facilities may also affect on the fuel cycle costs. In this paper a review is made of the future trend of nuclear fuel cycle cost of LWR based on the recent analysis. Those factors which affect the fuel cycle costs are also discussed. In order to reduce the uncertainties of the cost estimations as soon as possible, the necessity is emphasized to discuss internationally such items as the treatment and disposal of high level radioactive wastes, siting issues of a reprocessing plant, physical protection of plutonium and the effects of plutonium on the environment

Estimation of the core-wide fuel rod damage during a LWR LOCA

International Nuclear Information System (INIS)

Mattila, L.; Sairanen, R.; Stengaard, J.-O.

1975-01-01

The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

Integral Inherently Safe Light Water Reactor (I²S-LWR)

Energy Technology Data Exchange (ETDEWEB)

Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

2017-10-02

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I²S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

Least squares methodology applied to LWR-PV damage dosimetry, experience and expectations

International Nuclear Information System (INIS)

Wagschal, J.J.; Broadhead, B.L.; Maerker, R.E.

1979-01-01

The development of an advanced methodology for Light Water Reactors (LWR) Pressure Vessel (PV) damage dosimetry applications is the subject of an ongoing EPRI-sponsored research project at ORNL. This methodology includes a generalized least squares approach to a combination of data. The data include measured foil activations, evaluated cross sections and calculated fluxes. The uncertainties associated with the data as well as with the calculational methods are an essential component of this methodology. Activation measurements in two NBS benchmark neutron fields ( 252 Cf ISNF) and in a prototypic reactor field (Oak Ridge Pool Critical Assembly - PCA) are being analyzed using a generalized least squares method. The sensitivity of the results to the representation of the uncertainties (covariances) was carefully checked. Cross element covariances were found to be of utmost importance

Evaluating the loss of a LWR spent fuel or plutonium shipping package into the sea

International Nuclear Information System (INIS)

Heaberlin, S.W.; Baker, D.A.

1976-06-01

As the nations of the world turn to nuclear power for an energy source, commerce in nuclear fuel cycle materials will increase. Some of this commerce will be transported by sea. Such shipments give rise to the possibility of loss of these materials into the sea. This paper discusses the postulated accidental loss of two materials, light water reactor (LWR) spent fuel and plutonium, at sea. The losses considered are that of a single shipping package which is either undamaged or damaged by fire prior to the loss. The containment failure of the package in the sea,

Utilities/industries joint study on seismic isolation systems for LWR: Part I. Experimental and analytical studies on seismic isolation systems

International Nuclear Information System (INIS)

Kato, Muneaki; Sato, Shoji; Shimomura, Issei

1989-01-01

This paper describes a joint study program on seismic isolation systems for light-water reactors (LWRs) performed by ten electric power companies, three manufacturers, and five construction companies. The fundamental response characteristics of base-isolated structures and base-isolation devices are described. Applications of a base-isolation system to LWR buildings are given. Finally, three-dimensional shaking table experiments are described

The scale analysis sequence for LWR fuel depletion

International Nuclear Information System (INIS)

Hermann, O.W.; Parks, C.V.

1991-01-01

The SCALE (Standardized Computer Analyses for Licensing Evaluation) code system is used extensively to perform away-from-reactor safety analysis (particularly criticality safety, shielding, heat transfer analyses) for spent light water reactor (LWR) fuel. Spent fuel characteristics such as radiation sources, heat generation sources, and isotopic concentrations can be computed within SCALE using the SAS2 control module. A significantly enhanced version of the SAS2 control module, which is denoted as SAS2H, has been made available with the release of SCALE-4. For each time-dependent fuel composition, SAS2H performs one-dimensional (1-D) neutron transport analyses (via XSDRNPM-S) of the reactor fuel assembly using a two-part procedure with two separate unit-cell-lattice models. The cross sections derived from a transport analysis at each time step are used in a point-depletion computation (via ORIGEN-S) that produces the burnup-dependent fuel composition to be used in the next spectral calculation. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections. The techniques used by SAS2H and two recent applications of the code are reviewed in this paper. 17 refs., 5 figs., 5 tabs

Securing XML Documents

Directory of Open Access Journals (Sweden)

Charles Shoniregun

2004-11-01

Full Text Available XML (extensible markup language is becoming the current standard for establishing interoperability on the Web. XML data are self-descriptive and syntax-extensible; this makes it very suitable for representation and exchange of semi-structured data, and allows users to define new elements for their specific applications. As a result, the number of documents incorporating this standard is continuously increasing over the Web. The processing of XML documents may require a traversal of all document structure and therefore, the cost could be very high. A strong demand for a means of efficient and effective XML processing has posed a new challenge for the database world. This paper discusses a fast and efficient indexing technique for XML documents, and introduces the XML graph numbering scheme. It can be used for indexing and securing graph structure of XML documents. This technique provides an efficient method to speed up XML data processing. Furthermore, the paper explores the classification of existing methods impact of query processing, and indexing.

Evaluation of alternative waste management schemes for LWR hulls and caps

International Nuclear Information System (INIS)

Chaudon, L.; Cecille, L.; Klein, M.; Kowa, S.; Mehling, O.; Thiels, G.

1990-01-01

LWR hulls and caps represent one of the major sources of α bearing solid waste generated in the nuclear fuel cycle. For this reason, the CEC launched a theoretical study to evaluate alternative schemes for the overall management of this waste. Both volume reduction techniques and α decontamination of the hulls were assessed. The study demonstrated that the transport and disposal of the conditioned waste in deep geological formations play a dominant part in the total management costs. Important cost savings can be achieved through the implementation of efficient volume reduction techniques, i.e. melting or compaction. As an alternative approach, exhaustive α decontamination of the hulls appears promising, provided that the conditioned waste can be made to comply with the disposal criteria of mines. Finally, prolongation of the interim storage period for the waste packages from 1 to 30 years may prove beneficial on the transport costs

Supplmental design requirements document enhanced radioactive and mixed waste storage: Phase 5, Project W-113

International Nuclear Information System (INIS)