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


1

Nuclear Engineering Nuclear Criticality Safety  

E-Print Network [OSTI]

Nuclear Engineering Nuclear Criticality Safety The Nuclear Engineering Division (NE) of Argonne National Laboratory is experienced in performing criticality safety and shielding evaluations for nuclear, and neutron spectra. The NE nuclear criticality safety (NCS) capabilities are based on a staff with decades

Kemner, Ken

2

Nuclear Safety Regulatory Framework  

Broader source: Energy.gov (indexed) [DOE]

overall Nuclear Safety Policy & ESH Goals Safety Basis Review and Approval In the DOE governance model, contractors responsible for the facility develop the safety basis and...

3

Nuclear criticality safety guide  

SciTech Connect (OSTI)

This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

Pruvost, N.L.; Paxton, H.C. [eds.] [eds.

1996-09-01T23:59:59.000Z

4

Nuclear Safety Research and Development...  

Energy Savers [EERE]

Nuclear Safety Research and Development Proposal Review and Prioritization Process and Criteria Nuclear Safety Research and Development Program Office of Nuclear Safety Office of...

5

Nuclear Engineer (Criticality Safety)  

Broader source: Energy.gov [DOE]

This position is located in the Nuclear Safety Division (NSD) which has specific responsibility for managing the development, analysis, review, and approval of non-reactor nuclear facility safety...

6

General Engineer (Nuclear Safety)  

Broader source: Energy.gov [DOE]

The Chief of Nuclear Safety (CNS) reports the US/M&P; in serving as the Central Technical Authority (CTA) for M&P; activities, ensuring the Departments nuclear safety policies and...

7

Nuclear Explosive Safety Manual  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Manual provides supplemental details to support the requirements of DOE O 452.2D, Nuclear Explosive Safety.

2009-04-14T23:59:59.000Z

8

Nuclear Engineer (Nuclear Safety Specialist)  

Broader source: Energy.gov [DOE]

A successful candidate of this position will serve as a Nuclear Engineer (Nuclear Safety Specialist) responsible for day-to-day technical monitoring, and evaluation of aspects of authorization...

9

Nuclear Explosive Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, or successor directive, for routine and planned nuclear explosive operations (NEOs).

2015-01-26T23:59:59.000Z

10

Nuclear Explosive Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations (NEOs).

2014-07-10T23:59:59.000Z

11

Nuclear Safety (Pennsylvania)  

Broader source: Energy.gov [DOE]

The Nuclear Safety Division conducts a comprehensive nuclear power plant oversight review program of the nine reactors at the five nuclear power sites in Pennsylvania. It also monitors the...

12

Promulgating Nuclear Safety Requirements  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Applies to all Nuclear Safety Requirements Adopted by the Department to Govern the Conduct of its Nuclear Activities. Cancels DOE P 410.1. Canceled by DOE N 251.85.

1996-05-15T23:59:59.000Z

13

Nuclear Explosive Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Order establishes requirements to implement the nuclear explosive safety elements of DOE O 452.1D, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations. Cancels DOE O 452.2C. Admin Chg 1, 7-10-13

2009-04-14T23:59:59.000Z

14

Nuclear Explosive Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1D, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations (NEOs). Cancels DOE O 452.2C. Admin Chg 1, dated 7-10-13, cancels DOE O 452.2D.

2009-04-14T23:59:59.000Z

15

Nuclear Explosive Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The directive provides supplemental details to support the requirements of DOE O 452.2C, Nuclear Explosive Safety, dated 6-12-06. Canceled by DOE M 452.2-1A.

2006-06-12T23:59:59.000Z

16

Office of Nuclear Facility Safety Programs  

Broader source: Energy.gov [DOE]

The Office of Nuclear Facility Safety Programs establishes nuclear safety requirements related to safety management programs that are essential to the safety of DOE nuclear facilities.

17

Nuclear Explosive Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The directive establishes specific nuclear explosive safety (NES) program requirements to implement the DOE NES standards and other NES criteria for routine and planned nuclear explosive operations. Cancels DOE O 452.2B. Canceled by DOE O 452.2D.

2006-06-12T23:59:59.000Z

18

Nuclear Explosive Safety Manual - DOE Directives, Delegations...  

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

1A Admin Chg 1, Nuclear Explosive Safety Manual by Carl Sykes Functional areas: Administrative Change, Defense Nuclear Facility Safety and Health Requirement, Nuclear Safety,...

19

Nuclear Safety Information Dashboard | Department of Energy  

Office of Environmental Management (EM)

Nuclear Safety Information Dashboard Nuclear Safety Information Dashboard The Nuclear Safety Information (NSI) Dashboard provides a new user interface to the Occurrence Reporting...

20

Nuclear Safety News | Department of Energy  

Office of Environmental Management (EM)

Nuclear Safety News Nuclear Safety News October 4, 2012 Department of Energy Cites Battelle Energy Alliance, LLC for Nuclear Safety and Radiation Protection Violations The U.S....

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


21

Nuclear Explosive Safety Manual  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Department of Energy (DOE) Manual provides supplemental details on selected topics to support the requirements of DOE O 452.2D, Nuclear Explosive Safety, dated 4/14/09. Cancels DOE M 452.2-1. Admin Chg 1, dated 7-10-13, cancels DOE M 452.2-1A.

2009-04-14T23:59:59.000Z

22

Nuclear Explosive Safety Evaluation Processes  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Manual provides supplemental details to support the nuclear explosive safety evaluation requirement of DOE O 452.2D, Nuclear Explosive Safety. Does not cancel other directives. Admin Chg 1, 7-10-13.

2009-04-14T23:59:59.000Z

23

Nuclear reactor safety device  

DOE Patents [OSTI]

A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

24

CRAD, Facility Safety- Nuclear Facility Safety Basis  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Nuclear Facility Safety Basis.

25

Nuclear Safety and Global Cooperation.  

E-Print Network [OSTI]

??The thesis of is to strengthen the capacity building of nuclear safety and disaster prevention all over the world from a preventive perspective, and to… (more)

Chang, Yu-shan

2012-01-01T23:59:59.000Z

26

Safety Reports Series No. 11, Developing Safety Culture in Nuclear...  

Broader source: Energy.gov (indexed) [DOE]

in Nuclear Activities: Practical Suggestions to Assist Progress, International Atomic Energy Agency Safety Reports Series No. 11, Developing Safety Culture in Nuclear Activities:...

27

CRAD, Nuclear Safety Delegations for Documented Safety Analysis...  

Office of Environmental Management (EM)

Documented Safety Analysis Approval - January 8, 2015 (EA CRAD 31-09, Rev. 0) CRAD, Nuclear Safety Delegations for Documented Safety Analysis Approval - January 8, 2015 (EA CRAD...

28

Nuclear reactor safety device  

DOE Patents [OSTI]

A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

Hutter, E.

1983-08-15T23:59:59.000Z

29

Nuclear Explosive Safety Evaluation Processes  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Manual provides supplemental details to support the nuclear explosive safety (NES) evaluation requirement of Department of Energy (DOE) Order (O) 452.2D, Nuclear Explosive Safety, dated 4/14/09. Admin Chg 1, dated 7-10-13, cancels DOE M 452.2-2.

2009-04-14T23:59:59.000Z

30

Nuclear Reactor Safety Design Criteria  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Certified 11-18-10.

1993-01-19T23:59:59.000Z

31

Nuclear Explosive Safety Evaluation Processes - DOE Directives...  

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

2 Admin Chg 1, Nuclear Explosive Safety Evaluation Processes by Carl Sykes Functional areas: Administrative Change, Defense Nuclear Facility Safety and Health Requirement, Defense...

32

Independent Activity Report, Defense Nuclear Facilities Safety...  

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

Defense Nuclear Facilities Safety Board Public Meeting - October 2012 Independent Activity Report, Defense Nuclear Facilities Safety Board Public Meeting - October 2012 October...

33

Nuclear reactor safety heat transfer  

SciTech Connect (OSTI)

Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.

Jones, O.C.

1982-07-01T23:59:59.000Z

34

NRC - regulator of nuclear safety  

SciTech Connect (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) was formed in 1975 to regulate the various commercial and institutional uses of nuclear energy, including nuclear power plants. The agency succeeded the Atomic Energy Commission, which previously had responsibility for both developing and regulating nuclear activities. Federal research and development work for all energy sources, as well as nuclear weapons production, is now conducted by the U.S. Department of Energy. Under its responsibility to protect public health and safety, the NRC has three principal regulatory functions: (1) establish standards and regulations, (2) issue licenses for nuclear facilities and users of nuclear materials, and (3) inspect facilities and users of nuclear materials to ensure compliance with the requirements. These regulatory functions relate to both nuclear power plants and to other uses of nuclear materials - like nuclear medicine programs at hospitals, academic activities at educational institutions, research work, and such industrial applications as gauges and testing equipment. The NRC places a high priority on keeping the public informed of its work. The agency recognizes the interest of citizens in what it does through such activities as maintaining public document rooms across the country and holding public hearings, public meetings in local areas, and discussions with individuals and organizations.

NONE

1997-05-01T23:59:59.000Z

35

Safety of Nuclear Explosive Operations  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This directive establishes responsibilities and requirements to ensure the safety of routine and planned nuclear explosive operations and associated activities and facilities. Cancels DOE O 452.2A and DOE G 452.2A-1A. Canceled by DOE O 452.2C.

2001-08-07T23:59:59.000Z

36

Autoclave nuclear criticality safety analysis  

SciTech Connect (OSTI)

Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

1991-12-31T23:59:59.000Z

37

Total safety: A new safety culture to integrate nuclear safety and operational safety  

SciTech Connect (OSTI)

The creation of a complete and thorough safety culture is proposed for the purpose of providing additional assurance about nuclear safety and improving the performance of nuclear power plants. The safety philosophy developed a combination of the former hardware-oriented nuclear safety approach and recent operational safety concepts. The improvement of the latter, after TMI-2 and Chernobyl, has been proven very effective in reducing the total risk associated with nuclear power plants. The first part of this article introduces a {open_quotes}total safety{close_quotes} concept. This extends the concept of {open_quotes}nuclear safety{close_quotes} and makes it closer to the public perception of safety. This concept is defined by means of a taxonomy of total safety. The second part of the article shows that total safety can be achieved by integrating it into a modern quality assurance (QA) system since it is tailored to make implementation into a framework of QA easier. The author believes that the outstanding success experienced by various industries as a result of introducing the modern QA system should lead to its application for ensuring the safety and performance of nuclear facilities. 15 refs., 3 figs.

Saji, G. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan); Murphy, G.A. [ed.

1991-07-01T23:59:59.000Z

38

Elements of a nuclear criticality safety program  

SciTech Connect (OSTI)

Nuclear criticality safety programs throughout the United States are quite successful, as compared with other safety disciplines, at protecting life and property, especially when regarded as a developing safety function with no historical perspective for the cause and effect of process nuclear criticality accidents before 1943. The programs evolved through self-imposed and regulatory-imposed incentives. They are the products of conscientious individuals, supportive corporations, obliged regulators, and intervenors (political, public, and private). The maturing of nuclear criticality safety programs throughout the United States has been spasmodic, with stability provided by the volunteer standards efforts within the American Nuclear Society. This presentation provides the status, relative to current needs, for nuclear criticality safety program elements that address organization of and assignments for nuclear criticality safety program responsibilities; personnel qualifications; and analytical capabilities for the technical definition of critical, subcritical, safety and operating limits, and program quality assurance.

Hopper, C.M.

1995-07-01T23:59:59.000Z

39

Nuclear criticality safety: 2-day training course  

SciTech Connect (OSTI)

This compilation of notes is presented as a source reference for the criticality safety course. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used as Los Alamos; be able to identify examples of circumstances present during criticality accidents; have participated in conducting two critical experiments; be asked to complete a critique of the nuclear criticality safety training course.

Schlesser, J.A. [ed.] [comp.

1997-02-01T23:59:59.000Z

40

A philosophy for space nuclear systems safety  

SciTech Connect (OSTI)

The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The Nuclear Safety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions.

Marshall, A.C.

1992-08-01T23:59:59.000Z

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


41

Nuclear Safety Information Agreement Between the U.S. Nuclear...  

Office of Environmental Management (EM)

Operations (NRC)), Jim O'Brien, Director, Office of Nuclear Safety (EHSS DOE), Robert Johnson (Chief, Fuel Manufacturing Branch (NRC)) Front Row: Matt Moury, Associate Under...

42

WIPP-016, Rev. 0 Nuclear Criticality Safety Evaluation for  

E-Print Network [OSTI]

WIPP-016, Rev. 0 Nuclear Criticality Safety Evaluation for Contact-Handled Transuranic Waste/2008 Guidance (if applicable): _______________________ #12;NUCLEAR CRITICALITY SAFETY EVALUATION FOR CONTACT, directors, employees, agents, consultants or personal services contractors. #12;NUCLEAR CRITICALITY SAFETY

43

Nuclear Explosive Safety - DOE Directives, Delegations, and Requiremen...  

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

2D Admin Chg 1, Nuclear Explosive Safety by Carl Sykes This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of...

44

Nuclear Explosive Safety - DOE Directives, Delegations, and Requiremen...  

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

52.2E, Nuclear Explosive Safety by Angela Chambers Functional areas: Safety, Security This Department of Energy (DOE) Order establishes requirements to implement the nuclear...

45

Nuclear Safety Research and Development Annual Report, December...  

Energy Savers [EERE]

Nuclear Safety Research and Development Annual Report, December 2014 Nuclear Safety Research and Development Annual Report, December 2014 December 8, 2014 This document is the...

46

Nuclear Safety Research and Development Program Proposal Submittal...  

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

Nuclear Safety Research and Development Program Proposal Submittal Instructions for Fiscal Year 2015 1.0 INTRODUCTION The Nuclear Safety Research and Development (NSR&D) Program...

47

CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux...  

Broader source: Energy.gov (indexed) [DOE]

Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of...

48

Nuclear Safety Basis Program Review Overview and Management Oversight...  

Office of Environmental Management (EM)

Nuclear Safety Basis Program Review During Facility Operations and Transitions Volume 4 - Nuclear Safety Basis Program Review During Facility Decommissioning and Environmental...

49

Nuclear Safety at the Department of Energy  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Nuclear Safety is a core value of the Department of Energy. As our management principle state: "We will pursue our mission in a manner that is safe, secure, legally and ethically sound, and fiscally responsible."

2011-12-05T23:59:59.000Z

50

Nuclear Criticality Safety Application Guide: Safety Analysis Report Update Program  

SciTech Connect (OSTI)

Martin Marietta Energy Systems, Inc. (MMES) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Safety analyses are performed to identify hazards and potential accidents; to analyze the adequacy of measures taken to eliminate, control, or mitigate hazards; and to evaluate potential accidents and determine associated risks. Safety Analysis Reports (SARs) are prepared to document the safety analysis to ensure facilities can be operated safely and in accordance with regulations. Many of the facilities requiring a SAR process fissionable material creating the potential for a nuclear criticality accident. MMES has long had a nuclear criticality safety program that provides the technical support to fissionable material operations to ensure the safe processing and storage of fissionable materials. The guiding philosophy of the program has always been the application of the double-contingency principle, which states: {open_quotes}process designs shall incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.{close_quotes} At Energy Systems analyses have generally been maintained to document that no single normal or abnormal operating conditions that could reasonably be expected to occur can cause a nuclear criticality accident. This application guide provides a summary description of the MMES Nuclear Criticality Safety Program and the MMES Criticality Accident Alarm System requirements for inclusion in facility SARs. The guide also suggests a way to incorporate the analyses conducted pursuant to the double-contingency principle into the SAR. The prime objective is to minimize duplicative effort between the NCSA process and the SAR process and yet adequately describe the methodology utilized to prevent a nuclear criticality accident.

Not Available

1994-02-01T23:59:59.000Z

51

Office of Nuclear Safety Basis and Facility Design  

Broader source: Energy.gov [DOE]

The Office of Nuclear Safety Basis & Facility Design establishes safety basis and facility design requirements and expectations related to analysis and design of nuclear facilities to ensure protection of workers and the public from the hazards associated with nuclear operations.

52

DOE's Approach to Nuclear Facility Safety Analysis and Management  

Broader source: Energy.gov [DOE]

Presenter: Dr. James O'Brien, Director, Office of Nuclear Safety, Office of Health, Safety and Security, US Department of Energy

53

Nuclear safety information sharing agreement between NRC and...  

Office of Environmental Management (EM)

for DOE and NRC to exchange information related to safety issues associated with non-reactor nuclear facilities. The NRC-DOE Inter-Agency nuclear safety information sharing...

54

Management of National Nuclear Power Programs for assured safety  

SciTech Connect (OSTI)

Topics discussed in this report include: nuclear utility organization; before the Florida Public Service Commission in re: St. Lucie Unit No. 2 cost recovery; nuclear reliability improvement and safety operations; nuclear utility management; training of nuclear facility personnel; US experience in key areas of nuclear safety; the US Nuclear Regulatory Commission - function and process; regulatory considerations of the risk of nuclear power plants; overview of the processes of reliability and risk management; management significance of risk analysis; international and domestic institutional issues for peaceful nuclear uses; the role of the Institute of Nuclear Power Operations (INPO); and nuclear safety activities of the International Atomic Energy Agency (IAEA).

Connolly, T.J. (ed.)

1985-01-01T23:59:59.000Z

55

Safety of Decommissioning of Nuclear Facilities  

SciTech Connect (OSTI)

Full text of publication follows: ensuring safety during all stages of facility life cycle is a widely recognised responsibility of the operators, implemented under the supervision of the regulatory body and other competent authorities. As the majority of the facilities worldwide are still in operation or shutdown, there is no substantial experience in decommissioning and evaluation of safety during decommissioning in majority of Member States. The need for cooperation and exchange of experience and good practices on ensuring and evaluating safety of decommissioning was one of the outcomes of the Berlin conference in 2002. On this basis during the last three years IAEA initiated a number of international projects that can assist countries, in particular small countries with limited resources. The main IAEA international projects addressing safety during decommissioning are: (i) DeSa Project on Evaluation and Demonstration of Safety during Decommissioning; (ii) R{sup 2}D{sup 2}P project on Research Reactors Decommissioning Demonstration Project; and (iii) Project on Evaluation and Decommissioning of Former Facilities that used Radioactive Material in Iraq. This paper focuses on the DeSa Project activities on (i) development of a harmonised methodology for safety assessment for decommissioning; (ii) development of a procedure for review of safety assessments; (iii) development of recommendations on application of the graded approach to the performance and review of safety assessments; and (iv) application of the methodology and procedure to the selected real facilities with different complexities and hazard potentials (a nuclear power plant, a research reactor and a nuclear laboratory). The paper also outlines the DeSa Project outcomes and planned follow-up activities. It also summarises the main objectives and activities of the Iraq Project and introduces the R{sup 2}D{sup 2} Project, which is a subject of a complementary paper.

Batandjieva, B.; Warnecke, E.; Coates, R. [International Atomic Energy Agency, Vienna (Austria)

2008-01-15T23:59:59.000Z

56

CRAD, Nuclear Facility Safety System- September 25, 2009  

Broader source: Energy.gov [DOE]

Nuclear Facility Safety System Functionality Inspection Criteria, Inspection Activities, and Lines of Inquiry (HSS CRAD 64-17, Rev 0 )

57

Nuclear Safety | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission,

58

Ris-R-984(EN) Nuclear Safety Research  

E-Print Network [OSTI]

, Roskilde April 1997 #12;Abstract The report presents a summary of the work of the Nuclear Safety Re- search programmes and the Nordic Nuclear Safety Research Programme. This report describes the work of the departmentRisø-R-984(EN) Nuclear Safety Research and Facilities Department Annual Report 1996 Edited by B

59

Central Technical Authority Responsibilities Regarding Nuclear Safety Requirements  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The order establishes Central Technical Authority and Chief of Nuclear Safety/Chief of Defense Nuclear Safety responsibilities and requirements directed by the Secretary of Energy in the development and issuance of Department of Energy regulations and directives that affect nuclear safety. No cancellations.

2007-08-28T23:59:59.000Z

60

Nonreactor Nuclear Safety Design Criteria and Explosive Safety Criteria Guide for Use with DOE O 420.1, Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Guide provides guidance on the application of requirements for nonreactor nuclear facilities and explosives facilities of Department of Energy (DOE) O 420.1, Facility Safety, Section 4.1, Nuclear and Explosives Safety Design Criteria. No cancellation.

2000-03-28T23:59:59.000Z

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


61

Double-clad nuclear fuel safety rod  

DOE Patents [OSTI]

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

McCarthy, William H. (Los Altos, CA); Atcheson, Donald B. (Cupertino, CA); Vaidyanathan, Swaminathan (San Jose, CA)

1984-01-01T23:59:59.000Z

62

Safety Reports Series No. 11, Developing Safety Culture in Nuclear Activities: Practical Suggestions to Assist Progress, International Atomic Energy Agency  

Broader source: Energy.gov [DOE]

Safety Reports Series No. 11, Developing Safety Culture in Nuclear Activities: Practical Suggestions to Assist Progress, International Atomic Energy Agency

63

Safety program considerations for space nuclear reactor systems  

SciTech Connect (OSTI)

This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given.

Cropp, L.O.

1984-08-01T23:59:59.000Z

64

Ris-R-1104(EN) Nuclear Safety Research  

E-Print Network [OSTI]

.4 Radioanalytical chemistry 36 3.5 Ecophysiology 38 3.6 Radioactive waste 41 4 Nuclear facilities and services 43 4Risř-R-1104(EN) Nuclear Safety Research and Facilities Department Annual Report 1998 Edited by B, Roskilde, Denmark April 1999 #12;Abstract The report presents a summary of the work of the Nuclear Safety

65

Ris-M-2944(EN) Nuclear Safety Research  

E-Print Network [OSTI]

for Maintaining Nuclear Knowledge 23 3.3. DR 1 24 4. THE WASTE MANAGEMENT SECTION 26 4.1. Waste Management 26 4Risř-M-2944(EN) 3> Nuclear Safety Research Department Annual Progress Report 1990 Edited by F. Heikel Vinther Risř National Laboratory, Roskilde, Denmark July 1991 #12;Nuclear Safety Research

66

Ris-R-1162(EN) Nuclear Safety Research  

E-Print Network [OSTI]

.4 Radioanalytical chemistry 26 3.5 Radioactive waste 29 4 Nuclear facilities and services 30 4.1 Research reactor DRRisř-R-1162(EN) Nuclear Safety Research and Facilities Department Annual Report 1999 Edited by B, Denmark April 2000 #12;Abstract The report presents a summary of the work of the Nuclear Safety Research

67

Ris-R-679(EN) Nuclear Safety Research  

E-Print Network [OSTI]

of the nuclear facilities at Risø. The activities include personnel dosimetry, maintenance and calibra- tionRisø-R-679(EN) mil Nuclear Safety Research Department Annual Progress Report 1992 Edited by B March 1993 #12;Nuclear Safety Research K«*«i Department Annual Progress Report 1992 Edited by B

68

Ris-R-1318(EN) Nuclear Safety Research  

E-Print Network [OSTI]

. Risø Decommissioning has the task of preparing for the decommissioning of Risø's nuclear facilitiesRisø-R-1318(EN) Nuclear Safety Research Department Annual Report 2001 Edited by B. Majborn, A This report presents a summary of the work of the Nuclear Safety Research Department in 2001. The department

69

Ris-R-1019(EN) Nuclear Safety Research  

E-Print Network [OSTI]

.2 Severe accidents 7 2.3 Decommissioning of research reactors 9 2.4 Nuclear information 10 3 RadiationRisø-R-1019(EN) Nuclear Safety Research and Facilities Department Annual Report 1997 Edited by B of the work of the Nuclear Safety Research and Facilities Department in 1997. The department´s research

70

WIPP-025, Rev. 0 Summary of Nuclear Criticality Safety  

E-Print Network [OSTI]

at the Waste Isolation Pilot Plant #12;SUMMARY OF NUCLEAR CRITICALITY SAFETY EVALUATION FOR SHIELDED CONTAINERS PLANT WIPP-025, REV. 0 AUGUST 2009 Summary of Nuclear Criticality Safety Evaluation for Shielded ISOLATION PILOT PLANT WIPP-025, REV. 0 AUGUST 2009 ES-1 Executive Summary This report summarizes the nuclear

71

Nuclear Safety Workshop Summary | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAXBalanced ScorecardReactor Technology Subcommittee of NEACSummary Nuclear Safety

72

Nuclear Safety Policy - DOE Directives, Delegations, and Requirements  

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

ARCHIVED SEN-35-91, Nuclear Safety Policy by jnichols Functional areas: Environment, Safety, and Health, Canceled by DOE P 420.1 n3591.pdf -- PDF Document, 20 KB Writer: jnichols...

73

Safety Series No. 75-INSAG-4, Safety Culture: A report by the International Nuclear Safety Advisory Group, International Atomic Energy Agency  

Broader source: Energy.gov [DOE]

Safety Series No. 75-INSAG-4, Safety Culture: A report by the International Nuclear Safety Advisory Group, International Atomic Energy Agency, IAEA, 1991

74

Tutorial on nuclear thermal propulsion safety for Mars  

SciTech Connect (OSTI)

Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments.

Buden, D.

1992-01-01T23:59:59.000Z

75

Tutorial on nuclear thermal propulsion safety for Mars  

SciTech Connect (OSTI)

Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments.

Buden, D.

1992-08-01T23:59:59.000Z

76

Nuclear Facility Safety Basis Fundamentals Self-Study Guide Review...  

Broader source: Energy.gov (indexed) [DOE]

Oak Ridge Operations Nuclear Facility Safety Basis Fundamentals Self-Study Guide Review Questions Name: Organization: Directions: This is an open-book evaluation. Complete the...

77

Spent Nuclear Fuel project integrated safety management plan  

SciTech Connect (OSTI)

This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

Daschke, K.D.

1996-09-17T23:59:59.000Z

78

Nuclear Safety Enforcement Letter issued to Sandia Corporation...  

Broader source: Energy.gov (indexed) [DOE]

Enforcement Letter issued to Sandia Corporation Nuclear Safety Enforcement Letter issued to Sandia Corporation 9222014 Enforcement Letter, Sandia Corporation, September 22, 2014...

79

Senior Technical Safety Manager Qualification Program Self-Assessment- Chief of Nuclear Safety  

Broader source: Energy.gov [DOE]

This Chief of Nuclear Safety (CNS) Report was prepared to summarize the results of the July 2013 CNS self-assessment of the Senior Technical Safety Manager Qualification Program.

80

Ris-R-739(p*tf Nuclear Safety Research  

E-Print Network [OSTI]

D. i3 Risø-R-739(p*tf Nuclear Safety Research Department Annual Progress Report 1993 Edited by B 1994 #12;Nuclear Safety Research ****-mp, Risø, 1994 #12;Contents 1 latmfcKtioa 5 2 HcafchPhysics 5 2.1 Dosimetry 5 22 Development of Instruments

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


81

Ris-R-1232(EN) Nuclear Safety Research Department  

E-Print Network [OSTI]

Risø-R-1232(EN) Nuclear Safety Research Department Annual Report 2000 Edited by B. Majborn, S This report presents a summary of the work of the Nuclear Safety Re- search Department in 2000. The department for the tasks "Applied Health Physics and Emergency Preparedness", "Dosimetry", "Environmental Monitoring

82

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation  

SciTech Connect (OSTI)

One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

John D. Bess; J. Blair Briggs; David W. Nigg

2009-11-01T23:59:59.000Z

83

Nuclear Plant/Hydrogen Plant Safety: Issues and Approaches  

SciTech Connect (OSTI)

The U.S. Department of Energy, through its agents the Next Generation Nuclear Plant Project and the Nuclear Hydrogen Initiative, is working on developing the technologies to enable the large scale production of hydrogen using nuclear power. A very important consideration in the design of a co-located and connected nuclear plant/hydrogen plant facility is safety. This study provides an overview of the safety issues associated with a combined plant and discusses approaches for categorizing, quantifying, and addressing the safety risks.

Steven R. Sherman

2007-06-01T23:59:59.000Z

84

An Integrated Safety Assessment Methodology for Generation IV Nuclear Systems  

SciTech Connect (OSTI)

The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Early work of the RSWG focused on defining a safety philosophy founded on lessons learned from current and prior generations of nuclear technologies, and on identifying technology characteristics that may help achieve Generation IV safety goals. More recent RSWG work has focused on the definition of an integrated safety assessment methodology for evaluating the safety of Generation IV systems. The methodology, tentatively called ISAM, is an integrated “toolkit” consisting of analytical techniques that are available and matched to appropriate stages of Generation IV system concept development. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time.

Timothy J. Leahy

2010-06-01T23:59:59.000Z

85

Review of Nuclear Safety Culture at the Hanford Site Waste Treatment...  

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

Review of Nuclear Safety Culture at the Hanford Site Waste Treatment and Immobilization Plant Project, October 2010 Review of Nuclear Safety Culture at the Hanford Site Waste...

86

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

Not Available

1993-11-01T23:59:59.000Z

87

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

Not Available

1991-04-01T23:59:59.000Z

88

CRAD, New Nuclear Facility Documented Safety Analysis and Technical...  

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

Technical Safety Requirements - December 2, 2014 (EA CRAD 31-07, Rev. 0) More Documents & Publications CRAD, Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08...

89

Interface with the Defense Nuclear Facilities Safety Board  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The manual defines the process DOE will use to interface with the Defense Nuclear Facilities Safety Board and its staff. Canceled by DOE M 140.1-1A. Does not cancel other directives.

1996-12-30T23:59:59.000Z

90

Interface with the Defense Nuclear Facilities Safety Board  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Manual presents the process the Department of Energy will use to interface with the Defense Nuclear Facilities Safety Board (DNFSB) and its staff. Cancels DOE M 140.1-1A.

2001-03-30T23:59:59.000Z

91

Interface with the Defense Nuclear Facilities Safety Board  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Manual presents the process the Department of Energy will use to interface with the Defense Nuclear Facilities Safety Board (DNFSB) and its staff. Cancels DOE M 140.1-1.

1999-01-26T23:59:59.000Z

92

Nuclear Safety Workshop Agenda - Post Fukushima Initiatives and...  

Energy Savers [EERE]

Initiatives and Results In response to the March 2011 accident at the Fukushima Daiichi nuclear power plant, Secretary Chu initiated a series of actions to review the safety of...

93

Safety Oversight of Decommissioning Activities at DOE Nuclear Sites  

SciTech Connect (OSTI)

The Defense Nuclear Facilities Safety Board (Board) is an independent federal agency established by Congress in 1988 to provide nuclear safety oversight of activities at U.S. Department of Energy (DOE) defense nuclear facilities. The activities under the Board's jurisdiction include the design, construction, startup, operation, and decommissioning of defense nuclear facilities at DOE sites. This paper reviews the Board's safety oversight of decommissioning activities at DOE sites, identifies the safety problems observed, and discusses Board initiatives to improve the safety of decommissioning activities at DOE sites. The decommissioning of former defense nuclear facilities has reduced the risk of radioactive material contamination and exposure to the public and site workers. In general, efforts to perform decommissioning work at DOE defense nuclear sites have been successful, and contractors performing decommissioning work have a good safety record. Decommissioning activities have recently been completed at sites identified for closure, including the Rocky Flats Environmental Technology Site, the Fernald Closure Project, and the Miamisburg Closure Project (the Mound site). The Rocky Flats and Fernald sites, which produced plutonium parts and uranium materials for defense needs (respectively), have been turned into wildlife refuges. The Mound site, which performed R and D activities on nuclear materials, has been converted into an industrial and technology park called the Mound Advanced Technology Center. The DOE Office of Legacy Management is responsible for the long term stewardship of these former EM sites. The Board has reviewed many decommissioning activities, and noted that there are valuable lessons learned that can benefit both DOE and the contractor. As part of its ongoing safety oversight responsibilities, the Board and its staff will continue to review the safety of DOE and contractor decommissioning activities at DOE defense nuclear sites.

Zull, Lawrence M.; Yeniscavich, William [Defense Nuclear Facilities Safety Board, 625 Indiana Ave., NW, Suite 700, Washington, DC 20004-2901 (United States)

2008-01-15T23:59:59.000Z

94

THE IMPACT OF THE GLOBAL NUCLEAR SAFETY REGIME IN BRAZIL  

SciTech Connect (OSTI)

A turning point of the world nuclear industry with respect to safety occurred due to the accident at Chernobyl, in 1986. A side from the tragic personal losses and the enormous financial damage, the Chernobyl accident has literally demonstrated that ''a nuclear accident anywhere is an accident everywhere''. The impact was felt immediately by the nuclear industry, with plant cancellations (e.g. Austria), elimination of national programs (e.g. Italy) and general construction delays. However, the reaction of the nuclear industry was equally immediate, which led to the proposal and establishment of a Global Nuclear Safety Regime. This regime is composed of biding international safety conventions, globally accepted safety standard, and a voluntary peer review system. In a previous work, the author has presented in detail the components of this Regime, and briefly discussed its impact in the Brazilian nuclear power organizations, including the Regulatory Body. This work, on the opposite, briefly reviews the Global Nuclear Safety Regime, and concentrates in detail in the discussion of its impact in Brazil, showing how it has produced some changes, and where the peer pressure regime has failed to produce real results.

Almeida, C.

2004-10-06T23:59:59.000Z

95

Nonreactor Nuclear Safety Design Guide for use with DOE O 420.1C, Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Guide provides an acceptable approach for safety design of DOE hazard category 1, 2 and 3 nuclear facilities for satisfying the requirements of DOE O 420.1C. Cancels DOE G 420.1-1.

2012-12-04T23:59:59.000Z

96

Nonreactor Nuclear Safety Design Guide for use with DOE O 420...  

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

CURRENT DOE G 420.1-1A, Nonreactor Nuclear Safety Design Guide for use with DOE O 420.1C, Facility Safety by Pranab Guha Functional areas: Facility Safety, Nonreactor Nuclear...

97

FAQS Qualification Card – Nuclear Explosive Safety Study  

Broader source: Energy.gov [DOE]

A key element for the Department’s Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA).

98

Proceedings of the Nuclear Criticality Technology Safety Workshop  

SciTech Connect (OSTI)

This document contains summaries of most of the papers presented at the 1995 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 16 and 17 at San Diego, Ca. The meeting was broken up into seven sessions, which covered the following topics: (1) Criticality Safety of Project Sapphire; (2) Relevant Experiments For Criticality Safety; (3) Interactions with the Former Soviet Union; (4) Misapplications and Limitations of Monte Carlo Methods Directed Toward Criticality Safety Analyses; (5) Monte Carlo Vulnerabilities of Execution and Interpretation; (6) Monte Carlo Vulnerabilities of Representation; and (7) Benchmark Comparisons.

Rene G. Sanchez

1998-04-01T23:59:59.000Z

99

Nuclear Power - Operation, Safety and Environment  

E-Print Network [OSTI]

as operation, safety, environment and radiation effects. The book is not offering a comprehensive coverage of the material in each area. Instead, selected themes are highlighted by authors of individual chapters representing contemporary interests worldwide...

100

Aging of safety class 1E transformers in safety systems of nuclear power plants  

SciTech Connect (OSTI)

This report discusses aging effects on safety-related power transformers in nuclear power plants. It also evaluates maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission`s (NRC`s) Nuclear Plant-Aging Research approach. It investigates the materials used in transformer construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes transformer failure events reported in various databases, and evaluates maintenance practices. Databases maintained by the nuclear industry were analyzed to evaluate the effects of aging on the operation of nuclear power plants.

Roberts, E.W.; Edson, J.L.; Udy, A.C. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1996-02-01T23:59:59.000Z

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


101

Guidance for identifying, reporting and tracking nuclear safety noncompliances  

SciTech Connect (OSTI)

This document provides Department of Energy (DOE) contractors, subcontractors and suppliers with guidance in the effective use of DOE`s Price-Anderson nuclear safety Noncompliance Tracking System (NTS). Prompt contractor identification, reporting to DOE, and correction of nuclear safety noncompliances provides DOE with a basis to exercise enforcement discretion to mitigate civil penalties, and suspend the issuance of Notices of Violation for certain violations. Use of this reporting methodology is elective by contractors; however, this methodology is intended to reflect DOE`s philosophy on effective identification and reporting of nuclear safety noncompliances. To the extent that these expectations are met for particular noncompliances, DOE intends to appropriately exercise its enforcement discretion in considering whether, and to what extent, to undertake enforcement action.

NONE

1995-12-01T23:59:59.000Z

102

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Charges Relating to Nuclear Reactor Safety," 1976, availablestudies of light-water nuclear reactor safety, emphasizingstudies of overall nuclear reactor safety have been

Nero, A.V.

2010-01-01T23:59:59.000Z

103

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Charges Relating to Nuclear Reactor Safety," 1976, availableissues impor tant to nuclear reactor safety. This report wasstudies of overall nuclear reactor safety have been

Nero, A.V.

2010-01-01T23:59:59.000Z

104

Nuclear power safety in central and eastern Europe  

SciTech Connect (OSTI)

The Chernobyl accident showed the weaknesses in the Soviet approach to safety, particularly of nuclear reactors. Until recently, Western governments, scientists, and engineers did not understand how to help their Russian colleagues make a greater society. This article discusses the two main types of Soviet reactors, their safety problems, and the help Westerners are giving to make them safer. 35 refs., 1 fig., 4 tabs.

Wilson, R. [Harvard Univ., Cambridge, MA (United States)

1995-01-01T23:59:59.000Z

105

Systems Issues in Nuclear Reactor Safety  

E-Print Network [OSTI]

regulations 2 Traditional regulations Probabilistic Risk Assessment Risk-informed decision making Human-in-Depth is an element of the NRC's safety philosophy that employs successive compensatory measures 6 philosophy in the worst possible place. #12;Technological Risk Assessment (Reactors) · Study the system as an integrated

de Weck, Olivier L.

106

Department of Energy Nuclear Safety Policy  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

It is the policy of the Department of Energy to design, construct, operate, and decommission its nuclear facilities in a manner that ensures adequate protection of workers, the public, and the environment. Cancels SEN-35-91.

2011-02-08T23:59:59.000Z

107

Nuclear criticality safety tools in the Chernobyl-4 accident analysis  

SciTech Connect (OSTI)

The collaboration with the Italian Safety Authority (DISP), started in July 1986, has the aim of studying, from a neutronic point of view, the possible initiator event and the accident dynamics in unit four of the Chernobly nuclear power plant. This report was produced within the framework of that collaboration. A main condition of the present work was making use of standard calculational methods employed in nuclear criticality safety analysis. This means that the neutron multiplication factor calculation should be made with the modules and the cross-section libraries of the SCALE system or in any case with some KENO IV version and the burnup calculation with the ORIGEN code.

Landeyro, P.A.

1988-01-01T23:59:59.000Z

108

E-Print Network 3.0 - aerospace nuclear safety Sample Search...  

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

and Nuclear Technologies 2 A Systematic Approach to Safety Case Management Dr Tim Kelly Summary: The concept of the safety case' has already been adopted across many...

109

Office of Nuclear Safety | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGYWomen OwnedofDepartment ofJaredOakscience-based,OHAGasand FunctionheldNuclearNuclear

110

Nuclear space power safety and facility guidelines study  

SciTech Connect (OSTI)

This report addresses safety guidelines for space nuclear reactor power missions and was prepared by The Johns Hopkins University Applied Physics Laboratory (JHU/APL) under a Department of Energy grant, DE-FG01-94NE32180 dated 27 September 1994. This grant was based on a proposal submitted by the JHU/APL in response to an {open_quotes}Invitation for Proposals Designed to Support Federal Agencies and Commercial Interests in Meeting Special Power and Propulsion Needs for Future Space Missions{close_quotes}. The United States has not launched a nuclear reactor since SNAP 10A in April 1965 although many Radioisotope Thermoelectric Generators (RTGs) have been launched. An RTG powered system is planned for launch as part of the Cassini mission to Saturn in 1997. Recently the Ballistic Missile Defense Office (BMDO) sponsored the Nuclear Electric Propulsion Space Test Program (NEPSTP) which was to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source in space. As of late 1993 the flight portion of this program was canceled but work to investigate the attributes of the reactor were continued but at a reduced level. While the future of space nuclear power systems is uncertain there are potential space missions which would require space nuclear power systems. The differences between space nuclear power systems and RTG devices are sufficient that safety and facility requirements warrant a review in the context of the unique features of a space nuclear reactor power system.

Mehlman, W.F.

1995-09-11T23:59:59.000Z

111

Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities  

SciTech Connect (OSTI)

This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

Garvin, L.J.

1995-11-01T23:59:59.000Z

112

Proceedings of the Nuclear Criticality Technology and Safety Project Workshop  

SciTech Connect (OSTI)

This report is the proceedings of the annual Nuclear Criticality Technology and Safety Project (NCTSP) Workshop held in Monterey, California, on April 16--28, 1993. The NCTSP was sponsored by the Department of Energy and organized by the Los Alamos Critical Experiments Facility. The report is divided into six sections reflecting the sessions outlined on the workshop agenda.

Sanchez, R.G. [comp.

1994-01-01T23:59:59.000Z

113

Reducing nuclear danger through intergovernmental technical exchanges on nuclear materials safety management  

SciTech Connect (OSTI)

The United States and Russia are dismantling nuclear weapons and generating hundreds of tons of excess plutonium and high enriched uranium fissile nuclear materials that require disposition. The U.S. Department of Energy and Russian Minatom organizations.are planning and implementing safe, secure storage and disposition operations for these materials in numerous facilities. This provides a new opportunity for technical exchanges between Russian and Western scientists that can establish an improved and sustained common safety culture for handling these materials. An initiative that develops and uses personal relationships and joint projects among Russian and Western participants involved in fissile nuclear materials safety management contributes to improving nuclear materials nonproliferation and to making a safer world. Technical exchanges and workshops are being used to systematically identify opportunities in the nuclear fissile materials facilities to improve and ensure the safety of workers, the public, and the environment.

Jardine, L.J. [Lawrence Livermore National Lab., CA (United States); Peddicord, K.L. [Texas A and M Univ., College Station, TX (United States); Witmer, F.E.; Krumpe, P.F. [USDOE, Washington, DC (United States); Lazarev, L.; Moshkov, M. [Radievyj Inst., Leningrad (Russian Federation)

1997-04-09T23:59:59.000Z

114

SCALE 6: Comprehensive Nuclear Safety Analysis Code System  

SciTech Connect (OSTI)

Version 6 of the Standardized Computer Analyses for Licensing Evaluation (SCALE) computer software system developed at Oak Ridge National Laboratory, released in February 2009, contains significant new capabilities and data for nuclear safety analysis and marks an important update for this software package, which is used worldwide. This paper highlights the capabilities of the SCALE system, including continuous-energy flux calculations for processing multigroup problem-dependent cross sections, ENDF/B-VII continuous-energy and multigroup nuclear cross-section data, continuous-energy Monte Carlo criticality safety calculations, Monte Carlo radiation shielding analyses with automated three-dimensional variance reduction techniques, one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations, two- and three-dimensional lattice physics depletion analyses, fast and accurate source terms and decay heat calculations, automated burnup credit analyses with loading curve search, and integrated three-dimensional criticality accident alarm system analyses using coupled Monte Carlo criticality and shielding calculations.

Bowman, Stephen M [ORNL

2011-01-01T23:59:59.000Z

115

Floating nuclear power plant safety assurance principles  

SciTech Connect (OSTI)

In the north regions of the Russian federation and low density population areas, there is a real necessity for ecological clean energy small power sources. For this purpose, floating nuclear power plants, designed on the basis of atomic ship building engineering, are being conceptualized. It is possible to use the ship building plants for the reactor purposes. Issues such as radioactive waste management are described.

Zvonarev, B.M.; Kuchin, N.L.; Sergeev, I.V.

1993-12-31T23:59:59.000Z

116

Nuclear Safety Enforcement Documents | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014Nuclear

117

Risk Assessment in Support of DOE Nuclear Safety, Risk Information Notice, June 2010  

Broader source: Energy.gov [DOE]

On August 12, 2009, the Defense Nuclear Facilities Safety Board(DNFSB) issued Recommendation 2009?1, Risk Assessment Methodologies at Defense Nuclear Facilities. Thisrecommendation focused on the...

118

Software reliability and safety in nuclear reactor protection systems  

SciTech Connect (OSTI)

Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

1993-11-01T23:59:59.000Z

119

Reevaluating nuclear safety and security in a post 9/11 era.  

SciTech Connect (OSTI)

This report has the following topics: (1) Changing perspectives on nuclear safety and security; (2) Evolving needs in a post-9/11 era; (3) Nuclear Weapons--An attractive terrorist target; (4) The case for increased safety; (5) Evolution of current nuclear weapons safety and security; (6) Integrated surety; (7) The role of safety and security in enabling responsiveness; (8) Advances in surety technologies; and (9) Reevaluating safety.

Booker, Paul M.; Brown, Lisa M.

2005-07-01T23:59:59.000Z

120

Nuclear Criticality Safety | More Science | ORNL  

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

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


121

Nuclear safety | Princeton Plasma Physics Lab  

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

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122

Nuclear Safety Regulatory Framework | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission, Office of

123

Sandia National Laboratories: Nuclear Energy Safety Technologies  

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

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124

Princeton Plasma Physics Lab - Nuclear safety  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - SeptemberMicroneedles for4-16 FOR Primary Author Lastenergy Energy that

125

Q)Tf(^/7^,\\ Ris-R-625(pff Nuclear Safety Research  

E-Print Network [OSTI]

and nuclear releases. fields of radiation protection, reactor safety and radioactive waste managementQ)Tf(^/7^,\\ Risř-R-625(pff Nuclear Safety Research Department Annual Progress Report 1991 Edited Roskilde, Denmark March 1992 #12;Nuclear Safety Research Department Annual Progress Report 1991 Riso-R-62S

126

A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems  

E-Print Network [OSTI]

A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon through safety analy- sis is strongly mandated for safety-critical systems. Nuclear plant protection, NuFTA, for nuclear plant protection systems. NuFTA mechanically constructs a software fault tree

127

PLC-Based Safety Critical Software Development for Nuclear Power Plants  

E-Print Network [OSTI]

PLC-Based Safety Critical Software Development for Nuclear Power Plants Junbeom Yoo1 , Sungdeok Cha development technique for nuclear power plants'I&C soft- ware controllers. To improve software safety, we in developing safety-critical control software for a Korean nuclear power plant, and experience to date has been

128

Double-clad nuclear-fuel safety rod  

DOE Patents [OSTI]

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

McCarthy, W.H.; Atcheson, D.B.

1981-12-30T23:59:59.000Z

129

NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion  

SciTech Connect (OSTI)

An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top-level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of safety functional requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed. 9 refs.

Marshall, A.C.; Sawyer, J.C. Jr.; Bari, R.A.; Brown, N.W.; Cullingford, H.S.; Hardy, A.C.; Lee, J.H.; Mcculloch, W.H.; Niederauer, G.F.; Remp, K. (Sandia National Laboratories, Albuquerque, NM (United States) NASA, Washington (United States) Brookhaven National Laboratory, Upton, NY (United States) General Electric Co., San Jose, CA (United States) NASA, Johnson Space Center, Houston, Tn (United States) L

1992-07-01T23:59:59.000Z

130

NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion  

SciTech Connect (OSTI)

An Interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top- level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

Marshall, A.C.; Lee, J.H.; McCulloch, W.H. (Sandia National Labs., Albuquerque, NM (United States)); Sawyer, J.C. Jr. (National Aeronautics and Space Administration, Washington, DC (United States)); Bari, R.A. (Brookhaven National Lab., Upton, NY (United States)); Brown, N.W. (General Electric Co., San Jose, CA (United States)); Cullingford, H.S.; Hardy, A.C. (National Aeronautics and Space Administ

1992-01-01T23:59:59.000Z

131

NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion  

SciTech Connect (OSTI)

An Interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top- level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

Marshall, A.C.; Lee, J.H.; McCulloch, W.H. [Sandia National Labs., Albuquerque, NM (United States); Sawyer, J.C. Jr. [National Aeronautics and Space Administration, Washington, DC (United States); Bari, R.A. [Brookhaven National Lab., Upton, NY (United States); Brown, N.W. [General Electric Co., San Jose, CA (United States); Cullingford, H.S.; Hardy, A.C. [National Aeronautics and Space Administration, Houston, TX (United States). Lyndon B. Johnson Space Center; Niederauer, G.F. [Los Alamos National Lab., NM (United States); Remp, K. [National Aeronautics and Space Administration, Cleveland, OH (United States). Lewis Research Center; Rice, J.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Sholtis, J.A. [Department of the Air Force, Kirtland AFB, NM (United States)

1992-09-01T23:59:59.000Z

132

A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems  

E-Print Network [OSTI]

A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon tries to assure the systems' safety through performing various safety analysis techniques ­ FTA (Fault was KNICS(Korea Nuclear Instrumentation and Control System) RPS(Reactor Protection System). · Prototype

133

PRELIMINARY NUCLEAR CRITICALITY NUCLEAR SAFETY EVLAUATION FOR THE CONTAINER SURVEILLANCE AND STORAGE CAPABILITY PROJECT  

SciTech Connect (OSTI)

Washington Safety Management Solutions (WSMS) provides criticality safety services to Washington Savannah River Company (WSRC) at the Savannah River Site. One activity at SRS is the Container Surveillance and Storage Capability (CSSC) Project, which will perform surveillances on 3013 containers (hereafter referred to as 3013s) to verify that they meet the Department of Energy (DOE) Standard (STD) 3013 for plutonium storage. The project will handle quantities of material that are greater than ANS/ANSI-8.1 single parameter mass limits, and thus required a Nuclear Criticality Safety Evaluation (NCSE). The WSMS methodology for conducting an NCSE is outlined in the WSMS methods manual. The WSMS methods manual currently follows the requirements of DOE-O-420.1B, DOE-STD-3007-2007, and the Washington Savannah River Company (WSRC) SCD-3 manual. DOE-STD-3007-2007 describes how a NCSE should be performed, while DOE-O-420.1B outlines the requirements for a Criticality Safety Program (CSP). The WSRC SCD-3 manual implements DOE requirements and ANS standards. NCSEs do not address the Nuclear Criticality Safety (NCS) of non-reactor nuclear facilities that may be affected by overt or covert activities of sabotage, espionage, terrorism or other security malevolence. Events which are beyond the Design Basis Accidents (DBAs) are outside the scope of a double contingency analysis.

Low, M; Matthew02 Miller, M; Thomas Reilly, T

2007-04-30T23:59:59.000Z

134

Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report  

SciTech Connect (OSTI)

The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building).

MORGAN, R.G.

1999-02-23T23:59:59.000Z

135

Foundational development of an advanced nuclear reactor integrated safety code.  

SciTech Connect (OSTI)

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

2010-02-01T23:59:59.000Z

136

Implementing 10 CFR 830 at the FEMP Silos: Nuclear Health and Safety Plans as Documented Safety Analysis  

SciTech Connect (OSTI)

The objective of the Silos Project at the Fernald Closure Project (FCP) is to safely remediate high-grade uranium ore residues (Silos 1 and 2) and metal oxide residues (Silo 3). The evolution of Documented Safety Analyses (DSAs) for these facilities has reflected the changes in remediation processes. The final stage in silos DSAs is an interpretation of 10 CFR 830 Safe Harbor Requirements that combines a Health and Safety Plan with nuclear safety requirements. This paper will address the development of a Nuclear Health and Safety Plan, or N-HASP.

Fisk, Patricia; Rutherford, Lavon

2003-06-01T23:59:59.000Z

137

Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels  

E-Print Network [OSTI]

Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels B of Electronics and Computer Science, University of Southampton, Southampton, UK Abstract: Nuclear reactor and the usefulness of these robots for improving safety inspection of nuclear reactors in general are discussed

Chen, Sheng

138

ME 379M-Nuclear Safety and Security ABET EC2000 syllabus  

E-Print Network [OSTI]

ME 379M- Nuclear Safety and Security Page 1 ABET EC2000 syllabus ME 379M Nuclear Safety assessment models and nuclear non-proliferation. Failure classifications, failure modes, effects, and criticality analysis (FMECA), fault and event trees, reliability block diagrams. Specific areas from the code

Ben-Yakar, Adela

139

ROBERT J. BUDNITZ Occupation: Physicist in Energy/Environmental Research and Nuclear Safety  

E-Print Network [OSTI]

ROBERT J. BUDNITZ Occupation: Physicist in Energy/Environmental Research and Nuclear Safety Birth December 2004 to September 2007 (in Livermore): Leader, Nuclear & Risk Science Group, Energy & Environment Directorate Associate Program Leader for Nuclear Systems Safety and Security, E&E Directorate October 2002

Ajo-Franklin, Jonathan

140

Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications  

E-Print Network [OSTI]

C4.2 Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications and control (I&C) systems play a crucial role in the operation of nuclear power plants (NPP) and other safety is available. The use of model checking to verify two nuclear power plant related systems is described: an arc

Heljanko, Keijo

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


141

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.

1996-10-24T23:59:59.000Z

142

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.

1995-11-16T23:59:59.000Z

143

Reliability Engineering and System Safety 92 (2007) 609618 The nuclear industry's transition to risk-informed regulation and  

E-Print Network [OSTI]

Reliability Engineering and System Safety 92 (2007) 609­618 The nuclear industry's transition a Nuclear Science and Engineering Department, Massachusetts Institute of Technology, Cambridge, MA 02139, USA b Nuclear Power Engineering, Quality and Safety Management Department, Tokyo Electric Power

144

Office of Nuclear Safety Enforcement | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGYWomen OwnedofDepartment ofJaredOakscience-based,OHAGasand FunctionheldNuclear Safety

145

Pantex sets safety record | National Nuclear Security Administration  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006 TheSteven Ashby Dr. StevenPanoramicfirefighterssets safety record |

146

Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This standard provides a framework for generating Criticality Safety Evaluations (CSE) supporting fissionable material operations at Department of Energy (DOE) nonreactor nuclear facilities. This standard imposes no new criticality safety analysis requirements.

2007-02-07T23:59:59.000Z

147

Safety culture in the nuclear power industry : attributes for regulatory assessment  

E-Print Network [OSTI]

Safety culture refers to the attitudes, behaviors, and conditions that affect safety performance and often arises in discussions following incidents at nuclear power plants. As it involves both operational and management ...

Alexander, Erin L

2004-01-01T23:59:59.000Z

148

Nuclear criticality safety evaluation of Spray Booth Operations in X-705, Portsmouth Gaseous Diffusion Plant  

SciTech Connect (OSTI)

This report evaluates nuclear criticality safety for Spray Booth Operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current procedures and related hardware/equipment is presented. Control parameters relevant to nuclear criticality safety are explained, and a consolidated listing of administrative controls and safety systems is developed. Based on compliance with DOE Orders and MMES practices, the overall operation is evaluated, and recommendations for enhanced safety are suggested.

Sheaffer, M.K.; Keeton, S.C.

1993-09-20T23:59:59.000Z

149

Status and Value of International Standards for Nuclear Criticality Safety  

SciTech Connect (OSTI)

This presentation provides an update to the author's standards report provided at the ICNC-2007 meeting. It includes a discussion about the difference between, and the value of participating in, the development of international 'consensus' standards as opposed to nonconsensus standards. Standards are developed for a myriad of reasons. Generally, standards represent an agreed upon, repeatable way of doing something as defined by an individual or group of people. They come in various types. Examples include personal, family, business, industrial, commercial, and regulatory such as military, community, state, federal, and international standards. Typically, national and international 'consensus' standards are developed by individuals and organizations of diverse backgrounds representing the subject matter users and developers of a service or product and other interested parties or organizations. Within the International Organization for Standardization (ISO), Technical Committee 85 (TC85) on nuclear energy, Subcommittee 5 (SC5) on nuclear fuel technology, there is a Working Group 8 (WG8) on standardization of calculations, procedures, and practices related to criticality safety. WG8 has developed, and is developing, ISO standards within the category of nuclear criticality safety of fissionable materials outside of reactors (i.e., nonreactor fissionable material nuclear fuel cycle facilities). Since the presentation of the ICNC-2007 report, WG8 has issued three new finalized international standards and is developing two more new standards. Nearly all elements of the published WG8 ISO standards have been incorporated into IAEA nonconsensus guides and standards. The progression of consensus standards development among international partners in a collegial environment establishes a synergy of different concepts that broadens the perspectives of the members. This breadth of perspectives benefits the working group members in their considerations of consensus standards developments in their own countries. A testament to the value of the international standards efforts is that nearly all elements of the published WG8 ISO standards have been incorporated into IAEA nonconsensus guides and standards and are mainly consistent with international ISO member domestic standards.

Hopper, Calvin Mitchell [ORNL] [ORNL

2011-01-01T23:59:59.000Z

150

Underwater nuclear power plants: improved safety, environmental compatibility and efficiency  

SciTech Connect (OSTI)

The further development of nuclear power engineering depends on the creation of a new generation of nuclear power plant (NPP) projects that have a high degree of safety. Decisions ensuring secure NPP exploitation must be based on the possibility of eliminating or localizing accidents. Using environmental properties to achieve secure NPP exploitation and accident elimination leads to suggest the construction of NPPs in water. An efficient way to provide energy to remote coastal areas is through use of floatable construction of prefabricated units. Floatable construction raises the quality of works, reduces expenditures on industrial facilities, and facilities building conditions in districts with extreme climatic conditions. A type of NPP that is situated on a shelf with the reactor compartment placed at the sea bottom is proposed. The underwater location of the reactor compartment on the fixed depth allows the natural water environment conditions of natural hydrostatic pressure, heat transfer and circulation to provide NPP safety. An example of new concept for power units with under-water localization of the reactor compartment is provided by the double-block NPP in a VVER reactor.

Galustov, K.Z.; Abadjyan, K.A.; Pavlov, A.B.

1991-01-01T23:59:59.000Z

151

Nuclear Safety Risk Management in Refueling Outage of Qinshan Nuclear Power Plant  

SciTech Connect (OSTI)

The NPP is used to planning maintenance, in-service inspection, surveillance test, fuel handling and design modification in the refueling outage; the operator response capability will be reduced plus some of the plant systems out of service or loss of power at this time. Based on 8 times refueling outage experiences of the Qinshan NPP, this article provide some good practice and lesson learned for the nuclear safety risk management focus at four safety function areas of Residual Heat Removal Capability, Inventory Control, Power availability and Reactivity control. (authors)

Meijing Wu; Guozhang Shen [Qinshan Nuclear power company (China)

2006-07-01T23:59:59.000Z

152

Applications of nuclear data covariances to criticality safety and spent fuel characterization  

SciTech Connect (OSTI)

Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

Williams, Mark L [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Marshall, William BJ J [ORNL] [ORNL; Rearden, Bradley T [ORNL] [ORNL

2014-01-01T23:59:59.000Z

153

Cover letter, 10/29/03, re Nuclear Safety Technical Position, Deliverable 4.2.1  

Broader source: Energy.gov [DOE]

The enclosed Nuclear Safety Technical Position is Deliverable 4.2.1. under the Implementation Plan for Defense Nuclear Facilitises Board (DNFSB) Recommendation 2002-3, Requirements for Design...

154

Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities  

SciTech Connect (OSTI)

This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24.

Not Available

1993-11-01T23:59:59.000Z

155

Enforcement handbook: Enforcement of DOE nuclear safety requirements  

SciTech Connect (OSTI)

This Handbook provides detailed guidance and procedures to implement the General Statement of DOE Enforcement Policy (Enforcement Policy or Policy). A copy of this Enforcement Policy is included for ready reference in Appendix D. The guidance provided in this Handbook is qualified, however, by the admonishment to exercise discretion in determining the proper disposition of each potential enforcement action. As discussed in subsequent chapters, the Enforcement and Investigation Staff will apply a number of factors in assessing each potential enforcement situation. Enforcement sanctions are imposed in accordance with the Enforcement Policy for the purpose of promoting public and worker health and safety in the performance of activities at DOE facilities by DOE contractors (and their subcontractors and suppliers) who are indemnified under the Price-Anderson Amendments Act. These indemnified contractors, and their suppliers and subcontractors, will be referred to in this Handbook collectively as DOE contractors. It should be remembered that the purpose of the Department`s enforcement policy is to improve nuclear safety for the workers and the public, and this goal should be the prime consideration in exercising enforcement discretion.

NONE

1995-06-01T23:59:59.000Z

156

Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1  

SciTech Connect (OSTI)

Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

none,

1983-02-01T23:59:59.000Z

157

Safety Culture in Nuclear Installations | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Guidance for Use in the Enhancement of Safety Culture, International Atomic Energy Agency IAEA, December 2002. Developed for use in the IAEA's Safety Culture Services....

158

Nuclear Safety. Technical Progress Journal, October--December 1991: Volume 32, No. 4  

SciTech Connect (OSTI)

This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

Not Available

1991-01-01T23:59:59.000Z

159

ASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM-OF-SYSTEMS  

E-Print Network [OSTI]

by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe stateASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant

Paris-Sud XI, Université de

160

Myths and representations in French nuclear history: The impact on decommissioning safety  

E-Print Network [OSTI]

Myths and representations in French nuclear history: The impact on decommissioning safety C. Martin. The decommissioning of many operational plants (whether because nuclear power is being withdrawn, or because plants accident at the Fukushima nuclear power plant has shown that in many countries the debate on the withdrawal

Paris-Sud XI, Université de

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


161

International Symposium on Fusion Nuclear Technology (ISFNT-5) SAFETY ISSUES ASSOCIATED WITH MOBILIZED ACTIVATION  

E-Print Network [OSTI]

International Symposium on Fusion Nuclear Technology (ISFNT-5) SAFETY ISSUES ASSOCIATED;International Symposium on Fusion Nuclear Technology (ISFNT-5) heat from in-vessel systems with high neutron Symposium on Fusion Nuclear Technology (ISFNT-5) A design must adequately transfer heat from plasma

California at Los Angeles, University of

162

Nuclear criticality safety aspects of gaseous uranium hexafluoride (UF{sub 6}) in the diffusion cascade  

SciTech Connect (OSTI)

This paper determines the nuclear safety of gaseous UF{sub 6} in the current Gaseous Diffusion Cascade and auxiliary systems. The actual plant safety system settings for pressure trip points are used to determine the maximum amount of HF moderation in the process gas, as well as the corresponding atomic number densities. These inputs are used in KENO V.a criticality safety models which are sized to the actual plant equipment. The ENO V.a calculation results confirm nuclear safety of gaseous UF{sub 6} in plant operations..

Huffer, J.E. [Parallax, Inc., Atlanta, GA (United States)

1997-04-01T23:59:59.000Z

163

Office of Nuclear Safety and Environmental Assessments | Department...  

Energy Savers [EERE]

operation, deactivation, decontamination, decommissioning and environmental restoration. Conduct assessments of changes to operations, safety basis and modifications. Conducts...

164

Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant  

SciTech Connect (OSTI)

The Safety Technology Group is developing methodology that can be used to assess the risk of operating a plant to reprocess spent nuclear fuel. As an early step in the methodology, a preliminary hazards analysis identifies safety-related incidents. In the absence of appropriate safety features, these incidents could lead to significant consequences and risk to onsite personnel or to the public. This report is a compilation of potential safety-related incidents that have been identified in studies at SRL and in safety analyses of various commercially designed reprocessing plants. It is an expanded revision of the version originally published as DP-1558, Published December 1980.

Durant, W.S.; Perkins, W.C.; Lee, R.; Stoddard, D.H.

1982-05-20T23:59:59.000Z

165

Safety and Nonsafety Communications and Interactions in International Nuclear Power Plants  

SciTech Connect (OSTI)

Current industry and NRC guidance documents such as IEEE 7-4.3.2, Reg. Guide 1.152, and IEEE 603 do not sufficiently define a level of detail for evaluating interdivisional communications independence. The NRC seeks to establish criteria for safety systems communications that can be uniformly applied in evaluation of a variety of safety system designs. This report focuses strictly on communication issues related to data sent between safety systems and between safety and nonsafety systems. Further, the report does not provide design guidance for communication systems nor present detailed failure modes and effects analysis (FMEA) results for existing designs. This letter report describes communications between safety and nonsafety systems in nuclear power plants outside the United States. A limited study of international nuclear power plants was conducted to ascertain important communication implementations that might have bearing on systems proposed for licensing in the United States. This report provides that following information: 1.communications types and structures used in a representative set of international nuclear power reactors, and 2.communications issues derived from standards and other source documents relevant to safety and nonsafety communications. Topics that are discussed include the following: communication among redundant safety divisions, communications between safety divisions and nonsafety systems, control of safety equipment from a nonsafety workstation, and connection of nonsafety programming, maintenance, and test equipment to redundant safety divisions during operation. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.

Kisner, Roger A [ORNL; Mullens, James Allen [ORNL; Wilson, Thomas L [ORNL; Wood, Richard Thomas [ORNL; Korsah, Kofi [ORNL; Qualls, A L [ORNL; Muhlheim, Michael David [ORNL; Holcomb, David Eugene [ORNL; Loebl, Andy [ORNL

2007-08-01T23:59:59.000Z

166

Manual of functions, assignments, and responsibilities for nuclear safety: Revision 2  

SciTech Connect (OSTI)

The FAR Manual is a convenient easy-to-use collection of the functions, assignments, and responsibilities (FARs) of DOE nuclear safety personnel. Current DOE directives, including Orders, Secretary of Energy Notices, and other assorted policy memoranda, are the source of this information and form the basis of the FAR Manual. Today, the majority of FARs for DOE personnel are contained in DOE`s nuclear safety Orders. As these Orders are converted to rules in the Code of Federal Regulations, the FAR Manual will become the sole source for information relating to the functions, assignments, responsibilities of DOE nuclear safety personnel. The FAR Manual identifies DOE directives that relate to nuclear safety and the specific DOE personnel who are responsible for implementing them. The manual includes only FARs that have been extracted from active directives that have been approved in accordance with the procedures contained in DOE Order 1321.1B.

Not Available

1994-10-15T23:59:59.000Z

167

Improving the regulation of safety at DOE nuclear facilities. Final report  

SciTech Connect (OSTI)

The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by DOE itself. The three major recommendations are: under any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

NONE

1995-12-01T23:59:59.000Z

168

Improving the regulation of safety at DOE nuclear facilities. Final report: Appendices  

SciTech Connect (OSTI)

The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

NONE

1995-12-01T23:59:59.000Z

169

Nuclear Safety Component and Services Procurement, June 29, 2011...  

Office of Environmental Management (EM)

require component and materials replacement identified and implemented? * Are appropriate preventive maintenance requirements for stored safety-related equipment identified and...

170

Department of Energy Office of Nuclear Safety and Environmental...  

Broader source: Energy.gov (indexed) [DOE]

Safety, must comply with national consensus industry standards and the model building codes applicable for the state or region in which the facility is located. Certain...

171

Application of Neutron-Absorbing Structural-Amorphous metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls  

E-Print Network [OSTI]

241 Pu, etc. ). To prevent nuclear criticality in spent fuelto enhance criticality safety for spent nuclear fuel inSpent Nuclear Fuel (SNF) Container to Enhance Criticality

2006-01-01T23:59:59.000Z

172

CRAD, Nuclear Safety- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Nuclear Safety Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

173

Safety of nuclear power reactors in the former Eastern European countries  

SciTech Connect (OSTI)

This article discusses the safety of nuclear power plants in the former Eastern European countries (including the former Soviet Union). The current international design fabrication, construction, operation, safely, regulatory standards and practices, and ways to resolve plant problems are addressed in light of experience with the Western nuclear power development programs. 9 refs., 4 figs.

Chakraborty, S. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland)

1995-01-01T23:59:59.000Z

174

SHSD Manager Safety Engineering Group Manager  

E-Print Network [OSTI]

Safety, Machine Shop Safety, Tier I Program, Traffic Safety S. Moss: Nuclear Criticality Safety G. Shepherd: Explosives Safety, Facility Authorization Basis, Nuclear Safety R. Travis: Readiness Evaluations

175

WASTE PROCESSING ANNUAL NUCLEAR SAFETY RELATED R AND D REPORT FOR CY2008  

SciTech Connect (OSTI)

The Engineering and Technology Office of Waste Processing identifies and reduces engineering and technical risks associated with key waste processing project decisions. The risks, and actions taken to mitigate those risks, are determined through technology readiness assessments, program reviews, technology information exchanges, external technical reviews, technical assistance, and targeted technology development and deployment (TDD). The Office of Waste Processing TDD program prioritizes and approves research and development scopes of work that address nuclear safety related to processing of highly radioactive nuclear wastes. Thirteen of the thirty-five R&D approved work scopes in FY2009 relate directly to nuclear safety, and are presented in this report.

Fellinger, A.

2009-10-15T23:59:59.000Z

176

Safety of Department of Energy-Owned Nuclear Reactors  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establish reactor safety program requirements assure that the safety of each Department of Energy-owned (DOE-owned) reactor is properly analyzed, evaluated, documented, and approved by DOE; and reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate protection for health and safety and will be in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. Cancels Chap. 6 of DOE O 5480.1A. Paragraphs 7b(3), 7e(3) & 8c canceled by DOE O 5480.23 & canceled by DOE N 251.4 of 9-29-95.

1986-09-23T23:59:59.000Z

177

Preliminary nuclear safety assessment of the NEPST (Topaz II) space reactor program  

SciTech Connect (OSTI)

The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz II space nuclear power system. A preliminary nuclear safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclear safety assessment included a number of deterministic analyses, such as; neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz II system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

Marshall, A.C.

1993-01-01T23:59:59.000Z

178

Independent Oversight Assessment of the Nuclear Safety Culture...  

Office of Environmental Management (EM)

has also been effectively implemented in non-nuclear organizations, such as mining, health care, research, engineering, and transportation. The methodology entails collecting...

179

A new DOE standard for transuranic waste nuclear safety analysis  

SciTech Connect (OSTI)

The DOE Office of Environmental Management (EM) observed through onsite assessments and a review of site-specific lessons learned that transuranic (TRU) waste operations could benefit from standardization of assumptions and approaches used to analyze hazards and select controls. EM collected and compared safety analysis information from DOE sites, including a comparison of the type of TRU waste accidents evaluated and controls selected, as well as specific Airborne Release Fractions (ARFs), Respirable Fractions (RFs), and Damage Ratios (DRs) assumed in accident analyses. This paper recounts the efforts by the DOE and its contractors to bring consistency to the safety analysis process supporting TRU waste operations through an integrated re-engineering effort. EM embarked on a process to re-engineer and standardize TRU safety analysis activities complex-wide. The effort involved DOE headquarters, field offices, and contractors. Five teams were formed to analyze and develop the necessary technical basis for a DOE Technical Standard. The teams looked at general issues including Safety Basis (SB), drum integrity and inspection criteria, hazard controls and analysis, safety analysis review and approval process, and implementation of hazard controls. (authors)

Triay, I.; Chung, D. [U.S. Department of Energy, Washington, D.C. (United States); Woody, J. [Atlas Consulting, Knoxville, TN (United States); Foppe, T. [Carlsbad Technical Assistance Contractor, Carlsbad, NM (United States); Mewhinney, C. [Sandia National Laboratories, Carlsbad, NM (United States); Jennings, S. [Los Alamos National Laboratories, Carlsbad, NM (United States)

2007-07-01T23:59:59.000Z

180

Safety Software Guide Perspectives for the Design of New Nuclear Facilities (U)  

SciTech Connect (OSTI)

In June of this year, the Department of Energy (DOE) issued directives DOE O 414.1C and DOE G 414.1-4 to improve quality assurance programs, processes, and procedures among its safety contractors. Specifically, guidance entitled, ''Safety Software Guide for use with 10 CFR 830 Subpart A, Quality Assurance Requirements, and DOE O 414.1C, Quality Assurance, DOE G 414.1-4'', provides information and acceptable methods to comply with safety software quality assurance (SQA) requirements. The guidance provides a roadmap for meeting DOE O 414.1C, ''Quality Assurance'', and the quality assurance program (QAP) requirements of Title 10 Code of Federal Regulations (CFR) 830, Subpart A, Quality Assurance, for DOE nuclear facilities and software application activities. [1, 2] The order and guide are part of a comprehensive implementation plan that addresses issues and concerns documented in Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1. [3] Safety SQA requirements for DOE as well as National Nuclear Security Administration contractors are necessary to implement effective quality assurance (QA) processes and achieve safe nuclear facility operations. DOE G 414.1-4 was developed to provide guidance on establishing and implementing effective QA processes tied specifically to nuclear facility safety software applications. The Guide includes software application practices covered by appropriate national and international consensus standards and various processes currently in use at DOE facilities. While the safety software guidance is considered to be of sufficient rigor and depth to ensure acceptable reliability of safety software at all DOE nuclear facilities, new nuclear facilities are well suited to take advantage of the guide to ensure compliant programs and processes are implemented. Attributes such as the facility life-cycle stage and the hazardous nature of each facility operations are considered, along with the category and level of importance of the software. The discussion provided herein illustrates benefits of applying the Safety Software Guide to work activities dependent on software applications and directed toward the design of new nuclear facilities. In particular, the Guide-based systematic approach with software enables design processes to effectively proceed and reduce the likelihood of rework activities. Several application examples are provided for the new facility.

VINCENT, Andrew

2005-07-14T23:59:59.000Z

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


181

Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR  

SciTech Connect (OSTI)

This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

1983-08-01T23:59:59.000Z

182

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report  

SciTech Connect (OSTI)

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

183

Space nuclear safety program, May 1983. Progress report  

SciTech Connect (OSTI)

The studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems, pertained to the General-Purpose Heat Source (compatibility and safety verification) and to the Light-Weight Radioisotope Heater units (overpressure and impact tests).

Bronisz, S.E. (comp.)

1983-10-01T23:59:59.000Z

184

Aging of Class 1E batteries in safety systems of nuclear power plants  

SciTech Connect (OSTI)

This report presents the results of a study of aging effects on safety-related batteries in nuclear power plants. The purpose is to evaluate the aging effects caused by operation within a nuclear facility and to evaluate maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission's (NRC's) Nuclear Plant-Aging Research approach and investigates the materials used in battery construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes battery-failure events reported in various data bases, and evaluates recommended maintenance practices. Data bases that were analyzed included the NRC's Licensee Event Report system, the Institute for Nuclear Power Operations' Nuclear Plant Reliability Data System, the Oak Ridge National Laboratory's In-Plant Reliability Data System, and The S.M. Stoller Corporation's Nuclear Power Experience data base.

Edson, J.L.; Hardin, J.E.

1987-07-01T23:59:59.000Z

185

International Partnership on Mitigation and Measuring, Reporting...  

Open Energy Info (EERE)

for the Environment, Nature Conservancy and Nuclear Safety (BMU), German Agency for International Cooperation (GIZ) Sector Climate Focus Area Non-renewable Energy, Agriculture,...

186

South Africa-Developing Climate Policy Capacity within the South...  

Open Energy Info (EERE)

Affairs (DEA), Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) Sector Energy Focus Area Energy Efficiency Topics Background...

187

Thailand-Programme for Developing and Implementing a Climate...  

Open Energy Info (EERE)

Partner on behalf of the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) Sector Energy Focus Area Energy Efficiency Topics Low emission...

188

Democratic Republic of Congo-Low Emission Capacity Building Programme...  

Open Energy Info (EERE)

Programme (UNDP), German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU), Australian Department of Climate Change and Energy Efficiency...

189

Democratic Republic of Congo-EU-UNDP Low Emission Capacity Building...  

Open Energy Info (EERE)

Programme (UNDP), German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU), Australian Department of Climate Change and Energy Efficiency...

190

Trinidad and Tobago-EU-UNDP Low Emission Capacity Building Programme...  

Open Energy Info (EERE)

Programme (UNDP), German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU), Australian Department of Climate Change and Energy Efficiency...

191

Index to Nuclear Safety: a technical progress review by chronology, permuted title, and author, Volume 18 (1) through Volume 22 (6)  

SciTech Connect (OSTI)

This index to Nuclear Safety covers articles published in Nuclear Safety, Volume 18, Number 1 (January-February 1977) through Volume 22, Number 6 (November-December 1981). The index is divided into three section: a chronological list of articles (including abstracts), a permuted-title (KWIC) index, and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. Over 300 technical articles published in Nuclear Safety in the last 5 years are listed in this index.

Cottrell, W.B.; Passiakos, M.

1982-06-01T23:59:59.000Z

192

Nuclear criticality safety experiments, calculations, and analyses: 1958 to 1982. Volume 1. Lookup tables  

SciTech Connect (OSTI)

This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

Koponen, B.L.; Hampel, V.E.

1982-10-21T23:59:59.000Z

193

Nuclear criticality safety program at the University of Tennessee-Knoxville  

SciTech Connect (OSTI)

This paper presents an overview of the nuclear criticality safety (NCS) educational program at the University of Tennessee-Knoxville. The program is an academic specialization for nuclear engineering graduate students pursuing either the MS or PhD degree and includes special NCS courses and NCS research projects. Both the courses and the research projects serve as partial fulfillment of the requirements for the degree being pursued.

Basoglu, B.; Bentley, C.; Brewer, R.; Dunn, M.; Haught, C.; Plaster, M.; Wilkinson, A.; Dodds, H. (Univ. of Tennessee, Knoxville, TN (United States)); Elliott, E.; Waddell, W. (Martin Marietta Energy Systems Inc., Oak Ridge, TN (United States))

1993-01-01T23:59:59.000Z

194

Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant  

SciTech Connect (OSTI)

The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents.

Perkins, W.C.; Durant, W.S.; Dexter, A.H.

1980-12-01T23:59:59.000Z

195

Technical Basis for U. S. Department of Energy Nuclear Safety Policy, DOE Policy 420.1  

Broader source: Energy.gov [DOE]

This document provides the technical basis for the Department of Energy (DOE) Policy (P) 420.1, Nuclear Safety Policy, dated 2-8-2011. It includes an analysis of the revised Policy to determine whether it provides the necessary and sufficient high-level expectations that will lead DOE to establish and implement appropriate requirements to assure protection of the public, workers, and the environment from the hazards of DOE’s operation of nuclear facilities.

196

The impact of offsite factors on the safety performance of small nuclear power plants  

SciTech Connect (OSTI)

The results of an analysis of the influence of offsite factors on small nuclear power-plant (SNPP) safety performance during postulated severe accidents are presented. Given the plant locations in the immediate vicinity of residential areas and the impossibility of accomplishing the expeditious evacuation of the public, the risk caused by an SNPP severe accident may be considerably less than that for such an event in a large nuclear power plant. 3 refs., 3 figs., 5 tabs.

Baranaev, Yu.D.; Viktorov, A.N. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation)

1991-01-01T23:59:59.000Z

197

Space nuclear safety program. Progress report, July 1983  

SciTech Connect (OSTI)

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

Bronisz, S.E. (comp.)

1983-11-01T23:59:59.000Z

198

Superconducting Magnet Safety Nuclear Magnetic Resonance (NMR) facilities present unique hazards not found in most  

E-Print Network [OSTI]

Superconducting Magnet Safety Nuclear Magnetic Resonance (NMR) facilities present unique hazards not found in most laboratory environments. The NMR facilities maintain superconducting magnets which have for asphyxiation. Once energized the field of the superconducting magnet of the spectrometer is always present

Maroncelli, Mark

199

Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory  

SciTech Connect (OSTI)

One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.

J. D. Bess; J. B. Briggs; A. S. Garcia

2011-09-01T23:59:59.000Z

200

2012 Nuclear Safety Workshop Photos | Department of Energy  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustionImprovement Awardflash2007-42attachment1.pdfmodule(EE)2012 Nuclear Energy Enabling

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


201

2012 Nuclear Safety Workshop Presentations | Department of Energy  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustionImprovement Awardflash2007-42attachment1.pdfmodule(EE)2012 Nuclear Energy EnablingWorkshop

202

Assessment of safety-critical software in nuclear power plants  

SciTech Connect (OSTI)

This article outlines an approach in the design, documentation, and evaluation of computer systems. This allows the use of software in many safety-critical applications because it enables the systematic comparison of the program behavior with the engineering specifications of the computer system. Many of the ideas in this article have been used by the Atomic Energy Control Board of Canada in its safety assessment of the software for the shutdown systems of the Darlington Station. The four main elements of this approach follow: (1) Formal Documentation of Software Requirements: Most of the details of a complex environment can be ignored by system implementers and reviewers if they are given a complete and precise statement of the behavioral requirements for the computer system. We describe five mathematical relations that specify the requirements for the software in a computerized control system. (2) Design and Documentation of the Module Structure: Complexity caused by interactions between separately written components can be reduced by applying Data Abstraction, Abstract Data Types, and Object-Oriented Programming if the interfaces are precisely and completely documented. (3) Program Function Documentation: Software executions are lengthy sequences of state changes described by algorithms. The effects of these executive sequences can be precisely specified documented with tabular presentations of the program functions. Also, large programs can be decomposed and presented at a collection of well-documented smaller programs. (4) Tripod Approach to Assessment: There are three basic approaches to the assessment of complex software products: (i) testing, (ii) systematic inspection, and (iii) certification of people and processes. Assessment of a complex system cannot depend on any one of these alone. The approach used on the Darlington shutdown software, which included systematic inspection as well as planned and statistically designed random testing, is outlined.

Parnas, D.L.; Madey, J. [McMaster Univ., Hamilton, Ontario (Canada); Asmis, G.J.K. [Atomic Energy Control Board, Ottawa (Canada)

1991-04-01T23:59:59.000Z

203

Passive cooling safety system for liquid metal cooled nuclear reactors  

DOE Patents [OSTI]

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA); Hui, Marvin M. (Sunnyvale, CA); Berglund, Robert C. (Saratoga, CA)

1991-01-01T23:59:59.000Z

204

Nuclear Safety Research and Development Annual Report, December 2014 |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission, OfficeDepartment

205

Nuclear and Facility Safety Directives | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission,ScienceWasteand

206

Nuclear safety information sharing agreement between NRC and DOE's Office  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission,ScienceWasteandof

207

Office of Nuclear Safety and Environmental Assessments | Department of  

Broader source: Energy.gov (indexed) [DOE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGYWomen OwnedofDepartment ofJaredOakscience-based,OHAGasand FunctionheldNuclear

208

Preparation Of Nonreactor Nuclear Facility Documented Safety Analysis -  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - SeptemberMicroneedles for medical point07.06 Agenda 2006UptakeDOE Directives,

209

Pantex raises bike safety awareness | National Nuclear Security  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006 TheSteven Ashby Dr. StevenPanoramicfirefighters cook it

210

Pantex receives two safety awards | National Nuclear Security  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006 TheSteven Ashby Dr. StevenPanoramicfirefighters cook

211

Enforcement Regulations and Directives - Nuclear Safety | Department of  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in Review: TopEnergyIDIQBusinessinSupporting JobsCleanEnforcement2Energy Nuclear

212

NSC nears move completion with outstanding safety record | National Nuclear  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNational Nuclearhas 'Natitude' | National NuclearAdministrator|NewsSecuritySecurity

213

Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)  

SciTech Connect (OSTI)

The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses.

NONE

1993-09-01T23:59:59.000Z

214

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices  

SciTech Connect (OSTI)

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

215

ON THE USE OF SPRAY SYSTEMS: AN EXAMPLE OF R&D WORK IN HYDROGEN SAFETY FOR NUCLEAR APPLICATIONS  

E-Print Network [OSTI]

occurred since the Three Mile Island nuclear accident in 1979 through experimental programs1 ON THE USE OF SPRAY SYSTEMS: AN EXAMPLE OF R&D WORK IN HYDROGEN SAFETY FOR NUCLEAR APPLICATIONS, igniters and spray systems have been designed and installed in modern nuclear power plants. Mitigation

Boyer, Edmond

216

Review and Approval of Nuclear Facility Safety Basis and Safety Design  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0 Resource Program September 2010 B O N NReversing

217

Y-12's 1958 nuclear criticality accident and increased safety - 1958 brought accidents, more safety  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched FerromagnetismWaste andAnniversary, part 2 Continuing the 70 th Anniversary story

218

General-purpose heat source project and space nuclear safety fuels program. Progress report, February 1980  

SciTech Connect (OSTI)

This formal monthly report covers the studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are: General-Purpose Heat Source Development and Space Nuclear Safety and Fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work.

Maraman, W.J. (comp.)

1980-05-01T23:59:59.000Z

219

Implementation Evaluation Criteria for January 2001 Amended 10 CFR Part 830 Nuclear Safety Management  

SciTech Connect (OSTI)

This document provides criteria for use in performing gap evaluations of processes and documents relative to the requirements of 10 CFR Part 830, Nuclear Safety Management. The criteria and associated objective evidence statements have been approved by the cognizant interpretative authorities. The criteria have been developed for each section of 10 CFR Part 830. The criteria have been divided into two categories. Criteria and objective evidence have been developed for use in assessing Fluor Hanford (FH) programs and procedures at the company level--programmatic requirements and evidence. Criteria and objective evidence statements have also been developed for FH nuclear facilities and projects.

EVANS, C.B.

2001-02-13T23:59:59.000Z

220

Handbook of nuclear power plant seismic fragilities, Seismic Safety Margins Research Program  

SciTech Connect (OSTI)

The Seismic Safety Margins Research Program (SSMRP) has a gola to develop a complete fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor were made. These calculations required a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most would be usable for other sites as well.

Cover, L.E.; Bohn, M.P.; Campbell, R.D.; Wesley, D.A.

1983-12-01T23:59:59.000Z

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


221

Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel  

SciTech Connect (OSTI)

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k{sub eff}. Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

Marshall, William BJ J [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2012-01-01T23:59:59.000Z

222

Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

he purpose of this DOE Standard is to establish guidance for the preparation and review of hazard categorization and accident analyses techniques as required in DOE Order 5480.23, Nuclear Safety Analysis Reports.

1997-12-12T23:59:59.000Z

223

[6450-01-P], DEPARTMENT OF ENERGY, 10 CFR Part 830, Nuclear Safety Management, AGENCY: Department of Energy (DOE).  

Broader source: Energy.gov [DOE]

The Department of Energy (DOE) is issuing a final rule regarding Nuclear SafetyManagement. This Part establishes requirements for the safe management of DOE contractor andsubcontractor work at the...

224

The potential role of new technology for enhanced safety and performance of nuclear power plants through improved service maintenance  

E-Print Network [OSTI]

Refinements in the safety and performance of nuclear power plants must be made to maintain public confidence and ensure competitiveness with other power sources. The aircraft industry, US Navy, and other programs have ...

Achorn, Ted Glen

1991-01-01T23:59:59.000Z

225

Nuclear Safety Information Agreement Between the U.S. Nuclear Regulatory  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission, Office of Nuclear

226

Validation of Nuclear Criticality Safety Software and 27 energy group ENDF/B-IV cross sections  

SciTech Connect (OSTI)

The validation documented in this report is based on calculations that were executed during June through August 1992, and was completed in June 1993. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Martin Marietta Nuclear Criticality Safety Software. The first portion of the document outlines basic information in regard to validation of NCSS using ENDF/B-IV 27-group cross sections on the IBM 3090 at ORNL. A basic discussion of the NCSS system is provided, some discussion on the validation database and validation in general. Then follows a detailed description of the statistical analysis which was applied. The results of this validation indicate that the NCSS software may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant. When the validation results are treated as a single group, there is 95% confidence that 99.9% of future calculations of similar critical systems will have a calculated K{sub eff} > 0.9616. Based on this result the Portsmouth Nuclear Criticality Safety Department has adopted the calculational acceptance criteria that a k{sub eff} + 2{sigma} {le} 0.95 is safety subcritical. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25.

Lee, B.L. Jr.

1994-08-01T23:59:59.000Z

227

Techniques to evaluate the importance of common cause degradation on reliability and safety of nuclear weapons.  

SciTech Connect (OSTI)

As the nuclear weapon stockpile ages, there is increased concern about common degradation ultimately leading to common cause failure of multiple weapons that could significantly impact reliability or safety. Current acceptable limits for the reliability and safety of a weapon are based on upper limits on the probability of failure of an individual item, assuming that failures among items are independent. We expanded the current acceptable limits to apply to situations with common cause failure. Then, we developed a simple screening process to quickly assess the importance of observed common degradation for both reliability and safety to determine if further action is necessary. The screening process conservatively assumes that common degradation is common cause failure. For a population with between 100 and 5000 items we applied the screening process and conclude the following. In general, for a reliability requirement specified in the Military Characteristics (MCs) for a specific weapon system, common degradation is of concern if more than 100(1-x)% of the weapons are susceptible to common degradation, where x is the required reliability expressed as a fraction. Common degradation is of concern for the safety of a weapon subsystem if more than 0.1% of the population is susceptible to common degradation. Common degradation is of concern for the safety of a weapon component or overall weapon system if two or more components/weapons in the population are susceptible to degradation. Finally, we developed a technique for detailed evaluation of common degradation leading to common cause failure for situations that are determined to be of concern using the screening process. The detailed evaluation requires that best estimates of common cause and independent failure probabilities be produced. Using these techniques, observed common degradation can be evaluated for effects on reliability and safety.

Darby, John L.

2011-05-01T23:59:59.000Z

228

Nuclear Safety  

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

and Management Services Department at INL involves providing advanced risk and reliability analytical capabilities to support complex engineered facilities and processes. The...

229

Development of a Societal-Risk Goal for Nuclear Power Safety  

SciTech Connect (OSTI)

The safety-goal policy of the Nuclear Regulatory Commission (NRC) has never included a true societal-risk goal. The NRC did acknowledge that the original goal for the risk of latent cancer facilities “was an individual risk goal not related to the number of people involved,” and stated that “a true societal risk goal would place a limit on the aggregate number of people affected.” However, this limitation was never satisfactorily addressed. Moreover, the safety goal has historically focused primarily on fatalities and latent health effects, while experience with actual nuclear accidents has shown that societal disruption can be significant even in accidents that yield only small to modest numbers of fatalities. Therefore, we have evaluated the social disruption effects from severe reactor accidents as a basis to develop a societal-risk goal for nuclear power plants, considering both health effects and non-health concerns such as property damage and land interdiction. Our initial analysis considered six different nuclear power plant sites in the U.S. for Boiling Water Reactors and Pressurized Water Reactors. The accident sequences considered for these two reactor types were station blackout sequences (both short-term and long-term SBO) as well as an STSBO with RCIC failure for the BWR and a Steam Generator Tube Rupture for the PWR. The source term release was an input in a RASCAL calculation of the off-site consequences using actual site-based weather data for each of the six plant sites randomly selected over a two-year period. The source term release plumes were then compared to Geographical Information System data for each site to determine the population affected and that would need to be evacuated to meet current emergency preparedness regulations. Our results to date suggest that number of people evacuated to meet current protective action guidelines appears to be a good proxy for disruption -- and, unlike other measures of disruption, has the advantage of being relatively straightforward to calculate for a given accident scenario and a given geographical location and plant site. Revised safety goals taking into account the potential for societal disruption could in principle be applied to the current generation of nuclear plants, but could also be used in evaluating and siting new technologies, such as small modular light water reactors, advanced Gen-IV high-temperature reactors, as well as reactor designs with passive safety features such as filtered vented containments.

Vicki Bier; Michael Corradini; Robert Youngblood; Caleb Roh; Shuji Liu

2014-06-01T23:59:59.000Z

230

International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6) Nara, Japan, October 4-8, 2004.  

E-Print Network [OSTI]

for assuring quality of software. In the area of nuclear power plant control systems, testing on softwareThe 6th International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6) Nara, Japan, October 4-8, 2004. Paper ID. N6P298 Direct Control Flow Testing on Function Block Diagrams

231

Annual report to Congress. Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, calendar year 2000  

SciTech Connect (OSTI)

This Annual Report to the Congress describes the Department of Energy's activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board. During 2000, the Department completed its implementation and proposed closure of one Board recommendation and completed all implementation plan milestones associated with two additional Board recommendations. Also in 2000, the Department formally accepted two new Board recommendations and developed implementation plans in response to those recommendations. The Department also made significant progress with a number of broad-based safety initiatives. These include initial implementation of integrated safety management at field sites and within headquarters program offices, issuance of a nuclear safety rule, and continued progress on stabilizing excess nuclear materials to achieve significant risk reduction.

None

2001-03-01T23:59:59.000Z

232

Implementing Stakeholders' Access to Expertise: Experimenting on Nuclear Installations' Safety Cases - 12160  

SciTech Connect (OSTI)

In 2009 and 2010, the Institute for Nuclear Safety and Radiation Protection (IRSN) led two pilot actions dealing with nuclear installations' safety cases. One concerned the periodical review of the French 900 MWe nuclear reactors, the other concerned the decommissioning of a workshop located on the site of Areva's La Hague fuel-reprocessing plant site in Northwestern France. The purpose of both these programs was to test ways for IRSN and a small number of stakeholders (Non-Governmental Organizations (NGOs) members, local elected officials, etc.) to engage in technical discussions. The discussions were intended to enable the stakeholders to review future applications and provide valuable input. The test cases confirmed there is a definite challenge in successfully opening a meaningful dialogue to discuss technical issues, in particular the fact that most expertise reports were not public and the conflict that exists between the contrary demands of transparency and confidentiality of information. The test case also confirmed there are ways to further improvement of stakeholders' involvement. (authors)

Gilli, Ludivine; Charron, Sylvie [Institut de Radioprotection et de Surete Nucleaire (IRSN), Fontenay-aux-Roses (France)

2012-07-01T23:59:59.000Z

233

Institutional implications of establishing safety goals for nuclear power plants. [PWR; BWR  

SciTech Connect (OSTI)

The purpose of this project is to anticipate and address institutional problems that may arise from the adoption of NRC's proposed Policy Statement on Safety Goals for Nuclear Power Plants. The report emphasizes one particular category of institutional problems: the possible use of safety goals as a basis for legal challenges to NRC actions, and the resolution of such challenges by the courts. Three types of legal issues are identified and analyzed. These are, first, general legal issues such as access to the legal system, burden of proof, and standard of proof. Second is the particular formulation of goals. Involved here are such questions as sustainable rationale, definitions, avoided issues, vagueness of time and space details, and degree of conservatism. Implementation brings up the third set of issues which include interpretation and application, linkage to probabilistic risk assessment, consequences as compared to events, and the use of results.

Morris, F.A.; Hooper, R.L.

1983-07-01T23:59:59.000Z

234

Effect of {gamma}-irradiation on strength of concrete for nuclear-safety structures  

SciTech Connect (OSTI)

Concrete applied for construction of nuclear power plant (NPP) Temelin (Czech Republic) has been exposed to {gamma}-irradiation up to dose 6x10{sup 5} Gy. Depending on the level of irradiation, changes in strength, porous structure and phase composition of the concrete have been studied. It is found that irradiation lowers both the strength of concrete (about 10%) and volume (resp. surface) of porous space. On the other hand, {gamma}-irradiation increases the ratio of calcite, CaCO{sub 3}, in the concrete. Observed effects are discussed with respect to safety of NPPs.

Vodak, F. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic); Trtik, K. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic); Sopko, V. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic); Kapickova, O. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic); Demo, P. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic)]. E-mail: demo@fzu.cz

2005-07-01T23:59:59.000Z

235

Risk-Informing Safety Reviews for Non-Reactor Nuclear Facilities  

SciTech Connect (OSTI)

This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction help to risk-inform the safety review process and assess compliance with regulatory requirements.

Mubayi, V.; Azarm, A.; Yue, M.; Mukaddam, W.; Good, G.; Gonzalez, F.; Bari, R.A.

2011-03-13T23:59:59.000Z

236

Nuclear safety procedure upgrade project at USEC/MMUS gaseous diffusion plants  

SciTech Connect (OSTI)

Martin Marietta Utility Services has embarked on a program to upgrade procedures at both of its Gaseous Diffusion Plant sites. The transition from a U.S. Department of Energy government-operated facility to U.S. Nuclear Regulatory Commission (NRC) regulated has necessitated a complete upgrade of plant operating procedures and practices incorporating human factors as well as a philosophy change in their use. This program is designed to meet the requirements of the newly written 10CFR76, {open_quotes}The Certification of Gaseous Diffusion Plants,{close_quotes} and aid in progression toward NRC certification. A procedures upgrade will help ensure increased nuclear safety, enhance plant operation, and eliminate personnel procedure errors/occurrences.

Kocsis, F.J. III

1994-12-31T23:59:59.000Z

237

Monitoring the Long-Term Safety Performance of a Repository for Used Nuclear Fuel - 12294  

SciTech Connect (OSTI)

The nuclear waste management programs of several nations include plans for the design, construction and operation of deep geological repositories. Some of these programs have initiated the licensing process for their repository designs. Monitoring strategies and systems are at different levels of development in each program and there is common ground with respect to the ultimate goal of the monitoring function. In this context, the primary functions of a monitoring system are considered to be the verification of safety performance and making available information that may be required for implementation of future decisions such as the timing of repository decommissioning and closure or the possible retrieval of waste containers. This study examines some of the relevant issues and outlines a conceptual monitoring system for further study and development during implementation of Adaptive Phased Management, the method selected by the Government of Canada for long-term management of used nuclear fuel. (author)

Villagran, J.E. [Nuclear Waste Management Organization, Toronto (Canada)

2012-07-01T23:59:59.000Z

238

Operation Cornerstone onsite radiological safety report for announced nuclear tests, October 1988--September 1989  

SciTech Connect (OSTI)

Cornerstone was the name assigned to the series of underground nuclear experiments conducted at the Nevada Test Site (NTS) from October 1, 1988, through September 30, 1989. This report includes those experiments publicly announced. Remote radiation measurements were taken during and after each nuclear experiment by a telemetry system. Radiation Protection Technicians (RPT) with portable radiation detection instruments surveyed reentry routes into ground zeros (GZ) before other planned entries were made. Continuous surveillance was provided while personnel were in radiation areas and appropriate precautions were taken to protect persons from unnecessary exposure to radiation and toxic gases. Protective clothing and equipment were issued as needed. Complete radiological safety and industrial hygiene coverage were provided during drilling and mineback operations. Telemetered and portable radiation detector measurements are listed. Detection instrumentation used is described and specific operational procedures are defined.

Not Available

1990-08-01T23:59:59.000Z

239

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The objective of this Order is to establish facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. The Order has Change 1 dated 11-16-95, Change 2 dated 10-24-96, and the latest Change 3 dated 11-22-00 incorporated. The latest change satisfies a commitment made to the Defense Nuclear Facilities Safety Board (DNFSB) in response to DNFSB recommendation 97-2, Criticality Safety.

2000-11-20T23:59:59.000Z

240

Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants  

SciTech Connect (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) programs of applicants for nuclear power plant construction permits, operating licenses, standard design certifications, and combined operating licenses. The purpose of these safety reviews is to help ensure that personnel performance and reliability are appropriately supported. Detailed design review procedures and guidance for the evaluations is provided in three key documents: the Standard Review Plan (NUREG-0800), the HFE Program Review Model (NUREG-0711), and the Human-System Interface Design Review Guidelines (NUREG-0700). These documents were last revised in 2007, 2004 and 2002, respectively. The NRC is committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool. To this end, the NRC is updating its guidance to stay current with recent research on human performance, advances in HFE methods and tools, and new technology being employed in plant and control room design. This paper describes the role of HFE guidelines in the safety review process and the content of the key HFE guidelines used. Then we will present the methodology used to develop HFE guidance and update these documents, and describe the current status of the update program.

O, J.M.; Higgins, J.; Stephen Fleger - NRC

2011-09-19T23:59:59.000Z

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


241

BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities  

SciTech Connect (OSTI)

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. Data provided by these two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades The Russian Federation has been a major contributor to both projects with the Institute of Physics and Power Engineering (IPPE) as the major contributor from the Russian Federation. Included in the benchmark specifications from the BFS facilities are 34 critical configurations from BFS-49, 61, 62, 73, 79, 81, 97, 99, and 101; spectral characteristics measurements from BFS-31, 42, 57, 59, 61, 62, 73, 97, 99, and 101; reactivity effects measurements from BFS-62-3A; reactivity coefficients and kinetics measurements from BFS-73; and reaction rate measurements from BFS-42, 61, 62, 73, 97, 99, and 101.

J. Blair Briggs; Anatoly Tsibulya; Yevgeniy Rozhikhin

2012-03-01T23:59:59.000Z

242

Subject: Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities Project Number: 689Nuclear Energy Institute (NEI) Letter, 9/10/10  

Broader source: Energy.gov [DOE]

Enclosed for your review is a Nuclear Energy Institute white paper on the use of Integrated Safety Analysis (ISA) at U.S. Nuclear Regulatory Commission-licensed recycling facilities. This paper is...

243

Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society  

SciTech Connect (OSTI)

This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

Koponen, B.L.; Hampel, V.E.

1982-10-21T23:59:59.000Z

244

Annual report to Congress: Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, Calendar Year 1999  

SciTech Connect (OSTI)

This is the tenth Annual Report to the Congress describing Department of Energy activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board (Board). The Board, an independent executive-branch agency established in 1988, provides advice and recommendations to the Secretary of Energy regarding public health and safety issues at the Department's defense nuclear facilities. The Board also reviews and evaluates the content and implementation of health and safety standards, as well as other requirements, relating to the design, construction, operation, and decommissioning of the Department's defense nuclear facilities. During 1999, Departmental activities resulted in the closure of nine Board recommendations. In addition, the Department has completed all implementation plan milestones associated with three Board recommendations. One new Board recommendation was received and accepted by the Department in 1999, and a new implementation plan is being developed to address this recommendation. The Department has also made significant progress with a number of broad-based initiatives to improve safety. These include expanded implementation of integrated safety management at field sites, opening of a repository for long-term storage of transuranic wastes, and continued progress on stabilizing excess nuclear materials to achieve significant risk reduction.

None

2000-02-01T23:59:59.000Z

245

Annual report to Congress: Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, calendar year 1998  

SciTech Connect (OSTI)

This is the ninth Annual Report to the Congress describing Department of Energy (Department) activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board (Board). The Board, an independent executive-branch agency established in 1988, provides advice and recommendations to the Secretary of energy regarding public health and safety issues at the Department`s defense nuclear facilities. The Board also reviews and evaluates the content and implementation of health and safety standards, as well as other requirements, relating to the design, construction, operation, and decommissioning of the Department`s defense nuclear facilities. The locations of the major Department facilities are provided. During 1998, Departmental activities resulted in the proposed closure of one Board recommendation. In addition, the Department has completed all implementation plan milestones associated with four other Board recommendations. Two new Board recommendations were received and accepted by the Department in 1998, and two new implementation plans are being developed to address these recommendations. The Department has also made significant progress with a number of broad-based initiatives to improve safety. These include expanded implementation of integrated safety management at field sites, a renewed effort to increase the technical capabilities of the federal workforce, and a revised plan for stabilizing excess nuclear materials to achieve significant risk reduction.

NONE

1999-02-01T23:59:59.000Z

246

Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations  

SciTech Connect (OSTI)

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

247

Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization  

SciTech Connect (OSTI)

The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Robert Nourgaliev

2013-05-01T23:59:59.000Z

248

Consequence modeling for nuclear weapons probabilistic cost/benefit analyses of safety retrofits  

SciTech Connect (OSTI)

The consequence models used in former studies of costs and benefits of enhanced safety retrofits are considered for (1) fuel fires; (2) non-nuclear detonations; and, (3) unintended nuclear detonations. Estimates of consequences were made using a representative accident location, i.e., an assumed mixed suburban-rural site. We have explicitly quantified land- use impacts and human-health effects (e.g. , prompt fatalities, prompt injuries, latent cancer fatalities, low- levels of radiation exposure, and clean-up areas). Uncertainty in the wind direction is quantified and used in a Monte Carlo calculation to estimate a range of results for a fuel fire with uncertain respirable amounts of released Pu. We define a nuclear source term and discuss damage levels of concern. Ranges of damages are estimated by quantifying health impacts and property damages. We discuss our dispersal and prompt effects models in some detail. The models used to loft the Pu and fission products and their particle sizes are emphasized.

Harvey, T.F.; Peters, L.; Serduke, F.J.D.; Hall, C.; Stephens, D.R.

1998-01-01T23:59:59.000Z

249

Submersion Criticality Safety Analysis of Tungsten-Based Fuel for Nuclear Power and Propulsion Applications  

SciTech Connect (OSTI)

The Center for Space Nuclear Research (CSNR) is developing tungsten-encapsulated fuels for space nuclear applications. Aims to develop NTP fuels that are; Affordable Low impact on production and testing environment Producible on a large scale over suitable time period Higher-performance compared to previous graphite NTP fuel elements Space nuclear reactors remain subcritical before and during launch, and do not go critical until required by its mission. A properly designed reactor will remain subcritical in any launch abort scenario, where the reactor falls back to Earth and becomes submerged in terrestrial material. Submersion increases neutron reflection and thermalizes the neutrons, which typically increases the reactivity of the core. This effect is usually very significant for fast-spectrum reactors. This research provided a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor. Determine the submersion behavior of a reactor fueled by tungsten-based fuel. Considered fuel compositions with varying: Rhenium content (wt% rhenium in tungsten) Fuel loading fractions (UO2 vol%)

A.E. Craft; R. C. O'Brien; S. D. Howe; J. C. King

2014-07-01T23:59:59.000Z

250

The Development, Content, Design, and Conduct of the 2011 Piloted US DOE Nuclear Criticality Safety Program Criticality Safety Engineering Training and Education Project  

SciTech Connect (OSTI)

In May 1973 the University of New Mexico conducted the first nationwide criticality safety training and education week-long short course for nuclear criticality safety engineers. Subsequent to that course, the Los Alamos Critical Experiments Facility (LACEF) developed very successful 'hands-on' subcritical and critical training programs for operators, supervisors, and engineering staff. Since the inception of the US Department of Energy (DOE) Nuclear Criticality Technology and Safety Project (NCT&SP) in 1983, the DOE has stimulated contractor facilities and laboratories to collaborate in the furthering of nuclear criticality as a discipline. That effort included the education and training of nuclear criticality safety engineers (NCSEs). In 1985 a textbook was written that established a path toward formalizing education and training for NCSEs. Though the NCT&SP went through a brief hiatus from 1990 to 1992, other DOE-supported programs were evolving to the benefit of NCSE training and education. In 1993 the DOE established a Nuclear Criticality Safety Program (NCSP) and undertook a comprehensive development effort to expand the extant LACEF 'hands-on' course specifically for the education and training of NCSEs. That successful education and training was interrupted in 2006 for the closing of the LACEF and the accompanying movement of materials and critical experiment machines to the Nevada Test Site. Prior to that closing, the Lawrence Livermore National Laboratory (LLNL) was commissioned by the US DOE NCSP to establish an independent hands-on NCSE subcritical education and training course. The course provided an interim transition for the establishment of a reinvigorated and expanded two-week NCSE education and training program in 2011. The 2011 piloted two-week course was coordinated by the Oak Ridge National Laboratory (ORNL) and jointly conducted by the Los Alamos National Laboratory (LANL) classroom education and facility training, the Sandia National Laboratory (SNL) hands-on criticality experiments training, and the US DOE National Criticality Experiment Research Center (NCERC) hands-on criticality experiments training that is jointly supported by LLNL and LANL and located at the Nevada National Security Site (NNSS) This paper provides the description of the bases, content, and conduct of the piloted, and future US DOE NCSP Criticality Safety Engineer Training and Education Project.

Hopper, Calvin Mitchell [ORNL] [ORNL

2011-01-01T23:59:59.000Z

251

Safety research programs sponsored by Office of Nuclear Regulatory Research: Progress report, January 1--March 31, 1989  

SciTech Connect (OSTI)

This progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Regulatory Applications, Division of Engineering, Division of Safety Issue Resolution, and Division of Systems Research of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research following the reorganization in July 1988. The previous reports have covered the period October 1, 1976 through December 31, 1988.

Weiss, A.J. (comp.)

1989-08-01T23:59:59.000Z

252

The safety review and approval process for space nuclear power sources  

SciTech Connect (OSTI)

Over the past 30 yr. the U.S. Government has evolved a process for the safety review and launch approval of nuclear power sources (NPSs) proposed for launch into space. This process, which involves a number of governmental agencies, ensures that the various postulated accident scenarios are considered, that the responses of the NPSs to the accident environments are assessed, and that appropriate elements of the Federal Government are involved in the launch approval. This process has worked very well in the successful launches of 37 radioisotope thermoelectric generators and 1 reactor by the United States since 1961. Particular attention will be focused on the recent launch of the Galileo spacecraft. 19 refs., 12 figs., 4 tabs.

Bennett, G.L. [National Aeronautics and Space Administration, Washington, DC (United States)

1991-01-01T23:59:59.000Z

253

International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9) N9P0302 Kaohsiung, Taiwan, September 9-13, 2012  

E-Print Network [OSTI]

The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9 Kudinov Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, SE proportional to terminal melt spread thickness. At certain thickness, the melt layer becomes non

Haviland, David

254

Additional Studies of the Criticality Safety of Failed Used Nuclear Fuel  

SciTech Connect (OSTI)

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories and specific configurations were evaluated to understand trends and quantify the consequences of worst-case potential reconfiguration progressions. These results will be summarized here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g., >20% keff). It can be concluded that the consequences of credible fuel failure configurations from ES or transportation following ES are manageable (e.g., <5% keff). The current work expands on these efforts and examines some modified scenarios and modified approaches to investigate the effectiveness of some techniques for reducing the calculated increase in keff. The areas included here are more realistic modeling of some assembly types and the effect of reconfiguration of some assemblies in the storage and transportation canister.

Marshall, William BJ J [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2013-01-01T23:59:59.000Z

255

Aging of turbine drives for safety-related pumps in nuclear power plants  

SciTech Connect (OSTI)

This study was performed to examine the relationship between time-dependent degradation and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety-related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized-water reactor plants and in the Reactor Core Isolation Cooling and High-Pressure Coolant Injection systems for boiling-water reactor plants. This research has been conducted by examination of failure data in the Nuclear Plant Reliability Data System, review of Licensee Event Reports, discussion of problems with operating plant personnel, and personal observation. The reported failure data were reviewed to determine the cause of the event and the method of discovery. Based on the research results, attempts have been made to determine the predictability of failures and possible preventive measures that may be implemented. Findings in a recent study of AFW systems indicate that the turbine drive is the single largest contributor to AFW system degradation. However, examination of the data shows that the turbine itself is a reliable piece of equipment with a good service record. Most of the problems documented are the result of problems with the turbine controls and the mechanical overspeed trip mechanism; these apparently stem from three major causes which are discussed in the text. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design, have led to improved performance resulting in a reliable safety-related component. However, these improvements have not been universally implemented.

Cox, D.F. [Oak Ridge National Lab., TN (United States)

1995-06-01T23:59:59.000Z

256

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and the System Engineer Program.Chg 1 incorporates the use of DOE-STD-1189-2008, Integration of Safety into the Design Process, mandatory for Hazard Category 1, 2 and 3 nuclear facilities. Cancels DOE O 420.1A.

2005-12-22T23:59:59.000Z

257

Safety research programs sponsored by Office of Nuclear Regulatory Research: Quarterly progress report, July 1-September 30, 1986  

SciTech Connect (OSTI)

This progress report will describe current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Accident Evaluation, Division of Engineering Technology, and Division of Risk Analysis and Operations of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The projects reported are the following: High Temperature Reactor Research, SSC code improvements, Thermal-Hydraulic Reactor Safety Experiments, Thermodynamic Core-Concrete Interaction Experiments and Analysis, Plant Analyzer, Code Assessment and Application, Code Maintenance (RAMONA-3B), MELCOR Verification and Benchmarking, Source Term Code Package Verification and Benchmarking, Uncertainty Analysis of the Source Term; Stress Corrosion Cracking of PWR Steam Generator Tubing, Soil-Structure Interaction Evaluation and Structural Benchmarks, Identification of Age Related Failure Modes; Application of HRA/PRA Results to Support Resolution of Generic Safety Issues Involving Human Performance, Protective Action Decisionmaking, Rebaseling of Risk for Zion, Containment Performance Design Objective, and Operational Safety Reliability Research.

Bari, R.A.; Bezler, P.; Boccio, J.L.; Ginsberg, T.; Greene, G.A.; Guppy, J.G.; Hall, R.E.; Hofmayer, C.H.; Khatib-Rahbar, H.; Luckas, W.J. Jr.

1987-03-01T23:59:59.000Z

258

General-purpose heat source project and space nuclear safety and fuels program. Progress report  

SciTech Connect (OSTI)

Studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of LASL are presented. The three programs involved are: general-purpose heat source development; space nuclear safety; and fuels program. Three impact tests were conducted to evaluate the effects of a high temperature reentry pulse and the use of CBCF on impact performance. Additionally, two /sup 238/PuO/sub 2/ pellets were encapsulated in Ir-0.3% W for impact testing. Results of the clad development test and vent testing are noted. Results of the environmental tests are summarized. Progress on the Stirling isotope power systems test and the status of the improved MHW tests are indicated. The examination of the impact failure of the iridium shell of MHFT-65 at a fuel pass-through continued. A test plan was written for vibration testing of the assembled light-weight radioisotopic heater unit. Progress on fuel processing is reported.

Maraman, W.J.

1980-02-01T23:59:59.000Z

259

Stochastic sampling method with MCNPX for nuclear data uncertainty propagation in criticality safety applications  

SciTech Connect (OSTI)

In the domain of criticality safety, the efficient propagation of uncertainty in nuclear data to uncertainty in k{sub eff} is an important area of current research. In this paper, a method based on stochastic sampling is presented for uncertainty propagation in MCNPX calculations. To that aim, the nuclear data (i.e. cross sections) are assumed to have a multivariate normal distribution and simple random sampling is performed following this presumed probability distribution. A verification of the developed stochastic sampling procedure with MCNPX is then conducted using the {sup 239}Pu Jezebel experiment as well as the PB-2 BWR and TMI-1 PWR pin cell models from the Uncertainty Analysis in Modeling (UAM) exercises. For the Jezebel case, it is found that the developed stochastic sampling approach predicts similar k{sub eff} uncertainties compared to conventional sensitivity and uncertainty methods. For the UAM models, slightly lower uncertainties are obtained when comparing to existing preliminary results. Further details of these verification studies are discussed and directions for future work are outlined. (authors)

Zhu, T.; Vasiliev, A.; Wieselquist, W.; Ferroukhi, H. [Paul Scherrer Institut, 5232 Villigen (Switzerland)

2012-07-01T23:59:59.000Z

260

The ORSphere Benchmark Evaluation and Its Potential Impact on Nuclear Criticality Safety  

SciTech Connect (OSTI)

In the early 1970’s, critical experiments using an unreflected metal sphere of highly enriched uranium (HEU) were performed with the focus to provide a “very accurate description…as an ideal benchmark for calculational methods and cross-section data files.” Two near-critical configurations of the Oak Ridge Sphere (ORSphere) were evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results from those benchmark experiments were then compared with additional unmoderated and unreflected HEU metal benchmark experiment configurations currently found in the ICSBEP Handbook. For basic geometries (spheres, cylinders, and slabs) the eigenvalues calculated using MCNP5 and ENDF/B-VII.0 were within 3 of their respective benchmark values. There appears to be generally a good agreement between calculated and benchmark values for spherical and slab geometry systems. Cylindrical geometry configurations tended to calculate low, including more complex bare HEU metal systems containing cylinders. The ORSphere experiments do not calculate within their 1s uncertainty and there is a possibility that the effect of the measured uncertainties for the GODIVA I benchmark may need reevaluated. There is significant scatter in the calculations for the highly-correlated ORCEF cylinder experiments, which are constructed from close-fitting HEU discs and annuli. Selection of a nuclear data library can have a larger impact on calculated eigenvalue results than the variation found within calculations of a given experimental series, such as the ORCEF cylinders, using a single nuclear data set.

John D. Bess; Margaret A. Marshall; J. Blair Briggs

2013-10-01T23:59:59.000Z

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261

Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris  

DOE Patents [OSTI]

The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

Gabor, John D. (Western Springs, IL); Cassulo, John C. (Stickney, IL); Pedersen, Dean R. (Naperville, IL); Baker, Jr., Louis (Downers Grove, IL)

1986-01-01T23:59:59.000Z

262

Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants  

SciTech Connect (OSTI)

This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.

Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Hassan, M. [Brookhaven National Lab., Upton, NY (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States)

1998-01-01T23:59:59.000Z

263

Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors  

E-Print Network [OSTI]

L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

Galvez, Cristhian

2011-01-01T23:59:59.000Z

264

Expectations on Documented Safety Analysis for Deactivated Inactive Nuclear Facilities in a State of Long Term Surveillance & Maintenance or Decommissioning  

SciTech Connect (OSTI)

DOE promulgated 10 CFR 830 ''Nuclear Safety Management'' on October 10, 2000. Section 204 of the Rule requires that contractors at DOE hazard category 1, 2, and 3 nuclear facilities develop a ''Documented Safety Analysis'' (DSA) that summarizes the work to be performed, the associated hazards, and hazard controls necessary to protect workers, the public, and the environment. Table 2 of Appendix A to the rule has been provided to ensure that DSAs are prepared in accordance with one of the available predetermined ''safe harbor'' approaches. The table presents various acceptable safe harbor DSAs for different nuclear facility operations ranging from nuclear reactors to decommissioning activities. The safe harbor permitted for decommissioning of a nuclear facility encompasses methods described in DOE-STD-1 120-98, ''Integration of Environment, Safety and Health into Facility Disposition Activities,'' and provisions in 29 CFR 1910.120 or 29 CFR 1926.65 (HAZWOPER). Additionally, an evaluation of public safety impacts and development of necessary controls is required when the facility being decommissioned contains radiological inventory or contamination exceeding the Rule's definition for low-level residual fixed radioactivity. This document discusses a cost-effective DSA approach that is based on the concepts of DOE-STD-I 120 and meets the 10 CFR 830 safe harbor requirements for both transition surveillance and maintenance as well as decommissioning. This DSA approach provides continuity for inactive Hanford nuclear facilities that will eventually transition into decommissioning. It also uses a graded approach that meets the expectations of DOE-STD-3011 and addresses HAZWOPER requirements to provide a sound basis for worker protection, particularly where intrusive work is being conducted.

JACKSON, M.W.

2002-05-01T23:59:59.000Z

265

Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository--BMT1 of the DECOVALEX III project. Part 1: Conceptualization  

E-Print Network [OSTI]

Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste on the safety of a hypothetical nuclear waste repository at the near-field and are presented in three on the safety of nuclear waste repositories. To achieve the second objective, hypothetical benchmark test

Paris-Sud XI, Université de

266

Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics  

SciTech Connect (OSTI)

This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1993-03-01T23:59:59.000Z

267

U.S. Department of Energy, Oak Ridge Operations Office Nuclear Facility Safety Basis Fundamentals Self-Study Guide [Fulfills ORO Safety Basis Competency 1, 2 (Part 1), or 7 (Part 1)  

Broader source: Energy.gov [DOE]

"This self-study guide provides an overview of safety basis terminology, requirements, and activities that are applicable to DOE and Oak Ridge Operations Office (ORO) nuclear facilities on the Oak...

268

Facility Safety - DOE Directives, Delegations, and Requirements  

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

Change, Safety, The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety,...

269

Operational safety enhancement of Soviet-designed nuclear reactors via development of nuclear power plant simulators and transfer of related technology  

SciTech Connect (OSTI)

The US Department of Energy (DOE), under the US government`s International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper.

Kohut, P.; Epel, L.G.; Tutu, N.K. [and others

1998-08-01T23:59:59.000Z

270

Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code  

SciTech Connect (OSTI)

Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied more on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel-development activities. Realizing the full benefits of this approach will likely take some time. However, it is important that the developmental activities for modeling and simulation be tightly coupled with the experimental activities to maximize feedback effects and accelerate both the experimental and analytical elements of the program toward a common objective. The close integration of modeling and simulation and experimental activities is key to developing a useful fuel performance simulation capability, providing a validated design and analysis tool, and understanding the uncertainties within the models and design process. The efforts of this project are integrally connected to the Transmutation Fuels Campaign (TFC), which maintains as a primary objective to formulate, fabricate, and qualify a transuranic-based fuel with added minor actinides for use in future fast reactors. Additional details of the TFC scope can be found in the Transmutation Fuels Campaign Execution Plan. This project is an integral component of the TFC modeling and simulation effort, and this multiyear plan borrowed liberally from the Transmutation Fuels Campaign Modeling and Simulation Roadmap. This document provides the multiyear staged development plan to develop a continuum-level Integrated Performance and Safety Code (IPSC) to predict the behavior of the fuel and cladding during normal reactor operations and anticipated transients up to the point of clad breach.

Turner, John A [ORNL; Clarno, Kevin T [ORNL; Hansen, Glen A [ORNL

2009-09-01T23:59:59.000Z

271

New Resolved Resonance Region Evaluation for 63Cu and 65Cu for Nuclear Criticality Safety Program  

SciTech Connect (OSTI)

A new resolved resonance region evaluation of 63Cu and 65Cu was done in the energy region from 10-5 eV to 99.5 keV. The R-Matrix SAMMY method using the Reich-Moore approximation was used to create a new set of consistent resonance parameters. The new evaluation was based on three experimental transmission data sets; two measured at ORELA and one from MITR, and two radiative capture experimental data sets from GELINA. A total of 141 new resonances were identied for 63Cu and 117 for 65Cu. The corresponding set of external resonances for each isotope was based on the identied resonances above 99.5 keV from the ORELA transmission data. The negative external levels (bound levels) were determined to match the dierential thermal cross section measured at the MITR. Double dierential elastic scattering cross sections were calculated from the new set of resonance parameters. Benchmarking calculations were carried out on a set of ICSBEP benchmarks. This work is in support of the DOE Nuclear Criticality Safety Program.

Sobes, Vladimir [ORNL] [ORNL; Leal, Luiz C [ORNL] [ORNL; Guber, Klaus H [ORNL] [ORNL; Forget, Benoit [Massachusetts Institute of Technology (MIT)] [Massachusetts Institute of Technology (MIT); Kopecky, S. [EC-JRC-IRMM, Geel, Belgium] [EC-JRC-IRMM, Geel, Belgium; Schillebeeckx, P. [EC-JRC-IRMM, Geel, Belgium] [EC-JRC-IRMM, Geel, Belgium; Siegler, P. [EC-JRC-IRMM, Geel, Belgium] [EC-JRC-IRMM, Geel, Belgium

2014-01-01T23:59:59.000Z

272

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).  

SciTech Connect (OSTI)

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

Schultz, Peter Andrew

2011-12-01T23:59:59.000Z

273

Computer code for space-time diagnostics of nuclear safety parameters  

SciTech Connect (OSTI)

The computer code ECRAN 3D (Experimental and Calculation Reactor Analysis) is designed for continuous monitoring and diagnostics of reactor cores and databases for RBMK-1000 on the basis of analytical methods for the interrelation parameters of nuclear safety. The code algorithms are based on the analysis of deviations between the physically obtained figures and the results of neutron-physical and thermal-hydraulic calculations. Discrepancies between the measured and calculated signals are equivalent to obtaining inadequacy between performance of the physical device and its simulator. The diagnostics system can solve the following problems: identification of facts and time for inconsistent results, localization of failures, identification and quantification of the causes for inconsistencies. These problems can be effectively solved only when the computer code is working in a real-time mode. This leads to increasing requirements for a higher code performance. As false operations can lead to significant economic losses, the diagnostics system must be based on the certified software tools. POLARIS, version 4.2.1 is used for the neutron-physical calculation in the computer code ECRAN 3D. (authors)

Solovyev, D. A.; Semenov, A. A.; Gruzdov, F. V.; Druzhaev, A. A.; Shchukin, N. V.; Dolgenko, S. G.; Solovyeva, I. V.; Ovchinnikova, E. A. [National Research Nuclear Univ. MEPhI, Kashirskoe, 31, 115409, Moscow (Russian Federation)

2012-07-01T23:59:59.000Z

274

Aerosol particle transport modeling for preclosure safety studies of nuclear waste repositories  

SciTech Connect (OSTI)

An important concern for preclosure safety analysis of a nuclear waste repository is the potential release to the environment of respirable aerosol particles. Such particles, less than 10 {mu}m in aerodynamic diameter, may have significant adverse health effects if inhaled. To assess the potential health effects of these particles, it is not sufficient to determine the mass fraction of respirable aerosol. The chemical composition of the particles is also of importance since different radionuclides may pose vastly different health hazards. Thus, models are needed to determine under normal and accident conditions the particle size and the chemical composition distributions of aerosol particles as a function of time and of position in the repository. In this work a multicomponent sectional aerosol model is used to determine the aerosol particle size and composition distributions in the repository. A range of aerosol mass releases with varying mean particle sizes and chemical compositions is used to demonstrate the sensitivities and uncertainties of the model. Decontamination factors for some locations in the repository are presented. 8 refs., 1 tab.

Gelbard, F. [Sandia National Labs., Albuquerque, NM (USA)

1989-01-01T23:59:59.000Z

275

Assessing conformance to safety goals using nonparametric empirical Bayes methods: A nuclear reactor application  

SciTech Connect (OSTI)

Nonparametric empirical Bayes methods are used to develop decision criteria for use in deciding whether the risk of a given facility is compatible with a corresponding specified quantitative safety goal. The criteria utilize the uncertain results of a probabilistic risk assessment (PRA) and are derived from an empirical Bayes point of view in which the results from a set of similar facilities are used to estimate the population variability curve (PVC) for the parameter of interest. The PVC is estimated nonparametrically in the sense that the distributional family to which the PVC belongs is completely unknown and unspecified. For the assumed model, the method guarantees that all facilities ultimately accepted as being compatible with the goal have a prespecified exact assurance probability that the goal is not exceeded. The method also accounts for two possible biases in the PRA results. Criteria are developed for use in assessing the compatibility of nuclear power plant PRA-produced severe core damage frequency estimates with a corresponding subsidiary objective.

Martz, H.F.; Johnson, J.W. [Los Alamos National Lab., NM (United States)

1997-01-01T23:59:59.000Z

276

General-purpose heat source project and space nuclear safety and fuels program. Progress reportt, January 1980  

SciTech Connect (OSTI)

This formal monthly report covers the studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are the general-purpose heat source development and space nuclear safety and fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work.

Maraman, W.J. (comp.)

1980-04-01T23:59:59.000Z

277

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. Cancels DOE O 420.1B, DOE G 420.1-2 and DOE G 420.1-3.

2012-12-04T23:59:59.000Z

278

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. Cancels DOE 5480.7A, DOE 5480.24, DOE 5480.28 and Division 13 of DOE 6430.1A. Canceled by DOE O 420.1A.

1995-10-13T23:59:59.000Z

279

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Order establishes facility and programmatic safety requirements for Department of Energy facilities, which includes nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards mitigation, and the System Engineer Program. Cancels DOE O 420.1A. DOE O 420.1B Chg 1 issued 4-19-10.

2005-12-22T23:59:59.000Z

280

Nuclear Data for Criticality Safety and Reactor Applications at the Gaerttner LINAC Center Y. Danon, R.M. Bahran, E.J. Blain, A.M. Daskalakis, B.J. McDermott, D.G. Williams  

E-Print Network [OSTI]

Nuclear Data for Criticality Safety and Reactor Applications at the Gaerttner LINAC Center Y. Danon used in reactor and nuclear criticality safety applications. The goal of this program is to provide to nuclear criticality, neutron shielding applications, nuclear reactor design, and to better understand

Danon, Yaron

Note: This page contains sample records for the topic "nuclear safety bmu" from the National Library of EnergyBeta (NLEBeta).
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281

Facts and Lessons of the Fukushima Nuclear Accident and Safety Improvement- The Operator Viewpoints  

Broader source: Energy.gov [DOE]

Presenter: Akira Kawano, General Manager, Nuclear International Relations and Strategy Group, Nuclear Power and Plant Siting Administrative Department, Tokyo Electric Power Company

282

Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors  

E-Print Network [OSTI]

Advisory Committee and Generation IV International Forum.Nuclear Energy Agency The Generation IV International Forum.Technology Roadmap for Generation IV Nuclear Energy Systems.

Galvez, Cristhian

2011-01-01T23:59:59.000Z

283

Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process  

Broader source: Energy.gov [DOE]

Presenter: Undine Shoop, Chief, Health Physics and Human Performance Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission

284

Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques  

SciTech Connect (OSTI)

A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis and 2) topology-based methodologies to interactively visualize multidimensional data and extract risk-informed insights. Regarding item 1) we employ learning algorithms that aim to infer/predict simulation outcome and decide the coordinate in the input space of the next sample that maximize the amount of information that can be gained from it. Such methodologies can be used to both explore and exploit the input space. The later one is especially used for safety analysis scopes to focus samples along the limit surface, i.e. the boundaries in the input space between system failure and system success. Regarding item 2) we present a software tool that is designed to analyze multi-dimensional data. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations.

Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Diego Mandelli; Michael Pernice; Robert Nourgaliev

2013-10-01T23:59:59.000Z

285

ReseaRch at the University of Maryland Nuclear Safety Research at the University of Maryland  

E-Print Network [OSTI]

Research on nuclear energy started at the University of Maryland just after World War II, when and nuclear weapons was followed by controversial accidents and regulation. Today, nuclear power is considered that analyze the risks involved in the use of nuclear energy. Understanding and Using Radiation The ionizing

Hill, Wendell T.

286

Cognitive decision errors and organization vulnerabilities in nuclear power plant safety management: Modeling using the TOGA meta-theory framework  

SciTech Connect (OSTI)

In the field of nuclear power plant (NPP) safety modeling, the perception of the role of socio-cognitive engineering (SCE) is continuously increasing. Today, the focus is especially on the identification of human and organization decisional errors caused by operators and managers under high-risk conditions, as evident by analyzing reports on nuclear incidents occurred in the past. At present, the engineering and social safety requirements need to enlarge their domain of interest in such a way to include all possible losses generating events that could be the consequences of an abnormal state of a NPP. Socio-cognitive modeling of Integrated Nuclear Safety Management (INSM) using the TOGA meta-theory has been discussed during the ICCAP 2011 Conference. In this paper, more detailed aspects of the cognitive decision-making and its possible human errors and organizational vulnerability are presented. The formal TOGA-based network model for cognitive decision-making enables to indicate and analyze nodes and arcs in which plant operators and managers errors may appear. The TOGA's multi-level IPK (Information, Preferences, Knowledge) model of abstract intelligent agents (AIAs) is applied. In the NPP context, super-safety approach is also discussed, by taking under consideration unexpected events and managing them from a systemic perspective. As the nature of human errors depends on the specific properties of the decision-maker and the decisional context of operation, a classification of decision-making using IPK is suggested. Several types of initial situations of decision-making useful for the diagnosis of NPP operators and managers errors are considered. The developed models can be used as a basis for applications to NPP educational or engineering simulators to be used for training the NPP executive staff. (authors)

Cappelli, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Gadomski, A. M. [ECONA, Centro Interuniversitario Elaborazione Cognitiva Sistemi Naturali e Artificiali, via dei Marsi 47, Rome (Italy); Sepiellis, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Wronikowska, M. W. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Poznan School of Social Sciences (Poland)

2012-07-01T23:59:59.000Z

287

Nuclear reactor safety. Quarterly progress report, October 1-December 31, 1979  

SciTech Connect (OSTI)

Progress is reported in the following areas: LWRs, LMFBRs, HTGRs, GCFRs, and safety analysis of the TMI-2 severe overcooling accident. (DLC)

Jackson, J.F.; Stevenson, M.G. (comps.)

1980-05-01T23:59:59.000Z

288

Independent Oversight Assessment of the Nuclear Safety Culture and Management of Nuclear Safety Concerns at the Hanford Site Waste Treatment and Immobilization Plant, January 2012  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG | Department ofHTS CableDepartment ofDepartment2011 | DepartmentHealth, Safety

289

The safety climate of a Department of Energy nuclear facility: A sociotechnical analysis  

SciTech Connect (OSTI)

Government- and public-sponsored groups are increasingly demanding greater accountability by the Department of Energy`s weapons complex. Many of these demands have focused on the development of a positive safety climate, one that not only protects workers onsite, but also the surrounding populace and environment as well. These demands are, in part, a response to findings which demonstrate a close linkage between actual organizational safety performance and the organization`s safety climate, i.e., the collective attitudes employees hold concerning the level of safety in their organization. This paper describes the approach taken in the systematic assessment of the safety climate at EG&G Rocky Flats Plant (RFP).

Johnson, A.E.; Harbour, J.L.

1993-06-01T23:59:59.000Z

290

Review of nuclear power plant safety cable aging studies with recommendations for improved approaches and for future work.  

SciTech Connect (OSTI)

Many U. S. nuclear power plants are approaching 40 years of age and there is a desire to extend their life for up to 100 total years. Safety-related cables were originally qualified for nuclear power plant applications based on IEEE Standards that were published in 1974. The qualifications involved procedures to simulate 40 years of life under ambient power plant aging conditions followed by simulated loss of coolant accident (LOCA). Over the past 35 years or so, substantial efforts were devoted to determining whether the aging assumptions allowed by the original IEEE Standards could be improved upon. These studies led to better accelerated aging methods so that more confident 40-year lifetime predictions became available. Since there is now a desire to potentially extend the life of nuclear power plants way beyond the original 40 year life, there is an interest in reviewing and critiquing the current state-of-the-art in simulating cable aging. These are two of the goals of this report where the discussion is concentrated on the progress made over the past 15 years or so and highlights the most thorough and careful published studies. An additional goal of the report is to suggest work that might prove helpful in answering some of the questions and dealing with some of the issues that still remain with respect to simulating the aging and predicting the lifetimes of safety-related cable materials.

Gillen, Kenneth Todd; Bernstein, Robert

2010-11-01T23:59:59.000Z

291

Rewriting the standard on the functional requirements for computers used in safety systems of nuclear power plants  

SciTech Connect (OSTI)

Because of the rapid development of state-of-the-art computer technology, a rewrite of ANSI/IEEE-ANS-7-4.3.2-1982, {open_quotes}Application Criteria for Digital Computer Systems of Nuclear Power Generating Stations,{close_quotes} was required. This article outlines the thrust of this rewrite, which is nearing the balloting process, and identifies standards and guidelines to be used in the development of a highly reliable to be used in the development of a highly reliable computer system. The rewrite activity has been in process approximately 2 yr and is a cooperative project of the American Nuclear Society (ANS) Nuclear Power Plant Standards Committee (NUPPSCO) and the Institute of Electrical and Electronics Engineers (IEEE) Nuclear Power Engineering Committee (NPEC). Because computer technology has progressed significantly since ANSI/IEEE-ANS-7-4.3.2-1982 was issued, the rewrite was a very interesting challenge to the work group. The primary difference between the 1982 version and the rewrite is that the 1982 version addressed the quality assurance aspects of the Quality criteria, which included the integration of hardware and software and subsequent verification and validation, whereas the rewrite, being a product of IEEE Std 603-1991, {open_quotes}IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations,{close_quotes} was written to establish the minimum requirements for computer systems (hardware, software, and interfaces) used in safety systems. This article presents an introduction to the scope of the rewrite, provides a brief comment on how the work group addressed the scope, and follows with details on how the work group addressed the scope and rewrite direction. 1 fig., 1 tab.

Matras, J.R. [Science Applications International Corp., Las Vegas, NV (United States)

1991-07-01T23:59:59.000Z

292

Evaluation of station blackout accidents at nuclear power plants: Technical findings related to unresolved safety issue A-44: Final report  

SciTech Connect (OSTI)

''Station Blackout,'' which is the complete loss of alternating current (AC) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on AC power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed include: the fequency of loss of offsite power; the probability that emergency or onsite AC power supplies would be unavailable; the capability and reliability of decay heat removal systems independent of AC power; and the likelihood that offsite power would be restored before systems that cannot operate for extended periods without AC power fail, thus resulting in core damage. This report also addresses effects of different designs, locations, and operational features on the estimated frequency of core damage resulting from station blackout events.

Not Available

1988-06-01T23:59:59.000Z

293

An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility  

SciTech Connect (OSTI)

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

P. Calderoni; P. Sharpe; M. Shimada

2009-09-01T23:59:59.000Z

294

Spent nuclear fuel project cold vacuum drying facility safety equipment list  

SciTech Connect (OSTI)

This document provides the safety equipment list (SEL) for the Cold Vacuum Drying Facility (CVDF). The SEL was prepared in accordance with the procedure for safety structures, systems, and components (SSCs) in HNF-PRO-516, ''Safety Structures, Systems, and Components,'' Revision 0 and HNF-PRO-097, Engineering Design and Evaluation, Revision 0. The SEL was developed in conjunction with HNF-SO-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998). The SEL identifies the SSCs and their safety functions, the design basis accidents for which they are required to perform, the design criteria, codes and standards, and quality assurance requirements that are required for establishing the safety design basis of the SSCs. This SEL has been developed for the CVDF Phase 2 Safety Analysis Report (SAR) and shall be updated, expanded, and revised in accordance with future phases of the CVDF SAR until the CVDF final SAR is approved.

IRWIN, J.J.

1999-02-24T23:59:59.000Z

295

Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 14  

SciTech Connect (OSTI)

Supplement No. 14 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation with additional information submitted by the applicant since Supplement No. 13 was issued, and matters that the staff had under review when Supplement No. 13 was issued.

Tam, P.S.

1994-12-01T23:59:59.000Z

296

Guidelines for nuclear power plant safety issue prioritization information development. Supplement 4  

SciTech Connect (OSTI)

This is the fifth in a series of reports to document the use of a methodology developed by the Pacific Northwest Laboratory to calculate, for prioritization purposes, the risk, dose and cost impacts of implementing resolutions to reactor safety issues (NUREG/CR-2800, Andrews et al. 1983). This report contains results of issue-specific analyses for 23 issues. Each issue was considered within the constraints of available information as of winter 1986, and two staff-weeks of labor. The results are referenced, as one consideration in setting priorities for reactor safety issues, in NUREG-0933, ''A Prioritization of Generic Safety Issues.''

Tabatabai, A.S.; Fecht, B.A.; Powers, T.B.; Bickford, W.E.; Andrews, W.B.; Gallucci, R.H.V.; Bian, S.H.; Daling, P.M.; Eschbach, E.J.; Allen, C.H.

1986-07-01T23:59:59.000Z

297

Proposed risk evaluation guidelines for use by the DOE-AL Nuclear Explosive Safety Division in evaluating proposed shipments of nuclear components  

SciTech Connect (OSTI)

The licensing requirements of 10 CFR 71 (US Code of Federal Regulations) are the primary criteria used to license proposed US Department of Energy (DOE) shipments of nuclear components. However, if a shipment cannot meet 10 CFR 71 requirements, a Transportation System Risk Assessment (TSRA) is prepared to document: (1) the degree of compliance of proposed DOE shipments of nuclear components with applicable federal regulations, and (2) the risk associated with the proposed shipments. The Nuclear Explosive Safety Division (NESD) of the Department of Energy, Albuquerque Area Office (DOE-AL) is responsible for evaluating TSRAs and for preparing Safety Evaluation Reports (SERs) to authorize the off-site transport. Hazards associated with the transport may include the presence of fissile material, chemically and radiologically toxic uranium, and ionizing radiation. The Nuclear Regulatory Commission (NRC) has historically considered only radiological hazards in licensing the transport of radiological material because the US Department of Transportation considers licensing requirements of nonradiological (i.e., chemically toxic) hazards. The requirements of 10 CFR 71 are based primarily on consideration of radiological hazards. For completeness, this report provides information for assessing the effects of chemical toxicity. Evaluating the degree of compliance with the requirements of 10 CFR 71 is relatively straightforward. However, there are few precedents associated with developing TSRA risk assessments for packages that do not comply with all of the requirements of 10 CFR 71. The objective of the task is to develop Risk Evaluation Guidelines for DOE-AL to use when evaluating a TSRA. If the TSRA shows that the Risk Evaluation Guidelines are not exceeded, then from a risk perspective the TSRA should be approved if there is evidence that the ALARA (as low as reasonably achievable) principle has been applied.

Just, R.A.; Love, A.F.

1997-10-01T23:59:59.000Z

298

Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository--BMT1 of the DECOVALEX III project. Part 2: Effects of THM  

E-Print Network [OSTI]

Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste, Germany (7) Canadian Nuclear Safety Commission (CNSC), Ottawa, Canada (8) Lawrence Berkeley National Laboratory (LBNL), Berkeley, USA (9) INERIS-LAEGO, Ecole des Mines de Nancy, Nancy, France (10) Japan Nuclear

Paris-Sud XI, Université de

299

Safety Aspects of Wet Storage of Spent Nuclear Fuel, OAS-L-13-11  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartment of Energy fromCommentsRevolving STATEMENT OF Peter Malati, RahulSafetySafety

300

Summary and bibliography of safety-related events at pressurized-water nuclear power plants as reported in 1979  

SciTech Connect (OSTI)

This report summarizes the data contained in reports submitted by licensees to the US Nuclear Regulatory Commission concerning safety-related operational events that occurred at pressurized-water-reactor nuclear power plants in 1979. A bibliography containing 100-word abstracts of the event reports is included. The 2064 abstracts included in the bibliography describe incidents, failures, and design or construction deficiencies experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Keyword and permuted-title indexes are provided to facilitate location of the abstracts of interest. Tables summarizing the information contained in the bibliography are also presented and discussed. Information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and time of occurrence (i.e., during refueling, operation, testing, or construction). Some of the more interesting events that occurred during the year are reviewed in detail. 33 refs.

Scott, R.L.; Gallaher, R.B.

1981-07-01T23:59:59.000Z

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


301

Safety Bulletin  

Broader source: Energy.gov (indexed) [DOE]

in the documented safety analysis. BACKGROUND On March 11 , 2011 , the Fukushima Daiichi nuclear power station in Japan was damaged by a magnitude 9.0 earthquake and the...

302

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establish facility safety requirements for the Department of Energy, including National Nuclear Security Administration. Cancels DOE O 420.1. Canceled by DOE O 420.1B.

2002-05-20T23:59:59.000Z

303

Development of an Updated Societal-Risk Goal for Nuclear Power Safety  

SciTech Connect (OSTI)

This report briefly summarizes work done in FY 2013 on the subject LDRD. The working hypothesis is that societal disruption should be addressed in a safety goal. This is motivated by the point that the Fukushima disaster resulted in very little public dose, but enormous societal disruption; a goal that addressed societal disruption would fill a perceived gap in the US NRC safety goal structure. This year's work entailed analyzing the consequences of postulated accidents at various reactor sites in the US, specifically with a view to quantifying the number of people relocated and the duration of their relocation, to see whether this makes sense as a measure of societal disruption.

Vicki Bier; Michael Corradini; Robert Youngblood; Caleb Roh; Shuji Liu

2014-07-01T23:59:59.000Z

304

Review guidelines on software languages for use in nuclear power plant safety systems. Final report  

SciTech Connect (OSTI)

Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada, C/C++, Programmable Logic Controller (PLC) Ladder Logic, International Electrotechnical Commission (IEC) Standard 1131-3 Sequential Function Charts, Pascal, and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.s

Hecht, H.; Hecht, M.; Graff, S.; Green, W.; Lin, D.; Koch, S.; Tai, A.; Wendelboe, D. [SoHaR, Inc., Beverly Hills, CA (United States)

1996-06-01T23:59:59.000Z

305

Nuclear criticality safety bounding analysis for the in-tank-precipitation (ITP) process, impacted by fissile isotopic weight fractions  

SciTech Connect (OSTI)

The In-Tank Precipitation process (ITP) receives High Level Waste (HLW) supernatant liquid containing radionuclides in waste processing tank 48H. Sodium tetraphenylborate, NaTPB, and monosodium titanate (MST), NaTi{sub 2}O{sub 5}H, are added for removal of radioactive Cs and Sr, respectively. In addition to removal of radio-strontium, MST will also remove plutonium and uranium. The majority of the feed solutions to ITP will come from the dissolution of supernate that had been concentrated by evaporation to a crystallized salt form, commonly referred to as saltcake. The concern for criticality safety arises from the adsorption of U and Pt onto MST. If sufficient mass and optimum conditions are achieved then criticality is credible. The concentration of u and Pt from solution into the smaller volume of precipitate represents a concern for criticality. This report supplements WSRC-TR-93-171, Nuclear Criticality Safety Bounding Analysis For The In-Tank-Precipitation (ITP) Process. Criticality safety in ITP can be analyzed by two bounding conditions: (1) the minimum safe ratio of MST to fissionable material and (2) the maximum fissionable material adsorption capacity of the MST. Calculations have provided the first bounding condition and experimental analysis has established the second. This report combines these conditions with canyon facility data to evaluate the potential for criticality in the ITP process due to the adsorption of the fissionable material from solution. In addition, this report analyzes the potential impact of increased U loading onto MST. Results of this analysis demonstrate a greater safety margin for ITP operations than the previous analysis. This report further demonstrates that the potential for criticality in the ITP process due to adsorption of fissionable material by MST is not credible.

Bess, C.E.

1994-04-22T23:59:59.000Z

306

Stakeholder Transportation Scorecard: Reviewing Nevada's Recommendations for Enhancing the Safety and Security of Nuclear Waste Shipments - 13518  

SciTech Connect (OSTI)

As a primary stakeholder in the Yucca Mountain program, the state of Nevada has spent three decades examining and considering national policy regarding spent nuclear fuel and high-level radioactive waste transportation. During this time, Nevada has identified 10 issues it believes are critical to ensuring the safety and security of any spent nuclear fuel transportation program, and achieving public acceptance. These recommendations are: 1) Ship the oldest fuel first; 2) Ship mostly by rail; 3) Use dual-purpose (transportable storage) casks; 4) Use dedicated trains for rail shipments; 5) Implement a full-scale cask testing program; 6) Utilize a National Environmental Policy Act (NEPA) process for the selection of a new rail spur to the proposed repository site; 7) Implement the Western Interstate Energy Board (WIEB) 'straw man' process for route selection; 8) Implement Section 180C assistance to affected States, Tribes and localities through rulemaking; 9) Adopt safety and security regulatory enhancements proposed states; and 10) Address stakeholder concerns about terrorism and sabotage. This paper describes Nevada's proposals in detail and examines their current status. The paper describes the various forums and methods by which Nevada has presented its arguments and sought to influence national policy. As of 2012, most of Nevada's recommendations have been adopted in one form or another, although not yet implemented. If implemented in a future nuclear waste program, the State of Nevada believes these recommendations would form the basis for a successful national transportation plan for shipments to a geologic repository and/or centralized interim storage facility. (authors)

Dilger, Fred C. [Black Mountain Research, Henderson, NV 81012 (United States)] [Black Mountain Research, Henderson, NV 81012 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge, CA 91330 (United States)] [Department of Sociology, California State University, Northridge, CA 91330 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States)] [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States)

2013-07-01T23:59:59.000Z

307

Safety Software Quality Assurance Functions, Responsibilities, and Authorities for Nuclear Facilities and Activities  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To assign roles and responsibilities for improving the quality of safety software. DOE N 411.2 (archived) extends this Notice until 01/31/2005. DOE N 411.3 extends this Notice until 1/31/06. Canceled by DOE O 414.1C. does not cancel other directives.

2003-08-27T23:59:59.000Z

308

>r 9 5 6 O 0 V 3 Ris-R-807(EN) Nuclear Safety Research  

E-Print Network [OSTI]

. The activities cover hetlth physics, reactor physics, operation of the small reactor DR1, and radioactive waste, reactor safety and radioactive waste. In 1994 the department was organized in three sections: The Health. The Waste Management Section is responsible for the safe handling and storage of radioactive waste from Risø

309

2012 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety  

Broader source: Energy.gov [DOE]

Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

310

2011 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety  

Broader source: Energy.gov [DOE]

Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

311

2014 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety  

Broader source: Energy.gov [DOE]

Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

312

2013 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety  

Broader source: Energy.gov [DOE]

Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

313

Nuclear Fission Reactor Safety Research in FP7 and future perspectives  

E-Print Network [OSTI]

The European Union (?U) has defined in the Europe 2020 strategy and 2050 Energy Roadmap its long-term vision for establishing a secure, sustainable and competitive energy system and setting up legally binding targets by 2020 for reducing greenhouse emissions, by increasing energy efficiency and the share of renewable energy sources while including a significant share from nuclear fission. Nuclear energy can enable the further reduction in harmful emissions and can contribute to the EU’s competitive energy system, security of supply and independence from fossil fuels. Nuclear fission is a valuable option for those 14 EU countries that promote its use as part of their national energy mix. The European Group on Ethics in Science and New Technologies (EGE) adopted its Opinion No.27 ‘An ethical framework for assessing research, production and use of energy’ and proposed an integrated ethics approach for the research, production and use of energy in the EU, seeking equilibrium among four criteria – access ...

Garbil, Roger

2014-01-01T23:59:59.000Z

314

Aging and service wear of air-operated valves used in safety-related systems at nuclear power plants  

SciTech Connect (OSTI)

Air-operated valves (AOVs) are used in a variety of safety-related applications at nuclear power plants. They are often used where rapid stroke times are required or precise control of the valve obturator is required. They can be designed to operate automatically upon loss of power, which is often desirable when selecting components for response to design basis conditions. The purpose of this report is to examine the reported failures of AOVs and determine whether there are identifiable trends in the failures related to predictable causes. This report examines the specific components that comprise a typical AOV, how those components fail, when they fail, and how such failures are discovered. It also examines whether current testing frequencies and methods are effective in predicting such failures.

Cox, D.F.; McElhaney, K.L.; Staunton, R.H.

1995-05-01T23:59:59.000Z

315

Safety Series No. 75-INSAG-4, Safety Culture: A report by the...  

Broader source: Energy.gov (indexed) [DOE]

A report by the International Nuclear Safety Advisory Group, International Atomic Energy Agency Safety Series No. 75-INSAG-4, Safety Culture: A report by the International...

316

Nuclear Safety R&D in the Waste Processing Technology Development & Deployment Program  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September 15, 2010 PrintingNeed| Department ofDC.NavySafety R&D in the

317

An evaluation of the effectiveness of the US Department of Energy Integrated Safety Process (SS-21) for Nuclear Explosive Operations using quantitative hazard analysis  

SciTech Connect (OSTI)

This paper evaluates the effectiveness of the US Department of Energy Integrated Safety Process or ``Seamless Safety (SS-21)`` program for reducing risk associated with nuclear explosive operations. A key element in the Integrated Safety Process is the use of hazard assessment techniques to evaluate process design changes in parallel or concurrently with process design and development. This concurrent hazard assessment method recently was employed for the B61-0, 2 & 5 and W69 nuclear explosive dismantlement activities. This paper reviews the SS-21 hazard assessment process and summarizes the results of the concurrent hazard assessments performed for the B61 and W69 dismantlement programs. Comparisons of quantitative hazard assessment results before and after implementation of the SS-21 design process shed light on the effectiveness of the SS-21 program for achieving risk reduction.

Fischer, S.R.; Konkel, H.; Bott, T.; Eisenhawer, S.; Auflick, J.; Houghton, K.; Maloney, K.; DeYoung, L.; Wilson, M. [Los Alamos National Lab., NM (United States)]|[Sandia National Labs., Albuquerque, NM (United States)

1996-03-01T23:59:59.000Z

318

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

On other hand, accidents at nuclear facilities could nott ed expos ur e from a nuclear accident which would warrantresulting from accidents at nuclear facilities. Average

Nero, A.V.

2010-01-01T23:59:59.000Z

319

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

average value for nuclear plants) aFinal Envir. Statement (Statement, Koshkonong Nuclear Plant, August 1976. U. S.rem; operation of the nuclear plants themselves only *Other

Nero, A.V.

2010-01-01T23:59:59.000Z

320

Application of Neutron-Absorbing Structural-Amorphous Metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Control  

SciTech Connect (OSTI)

This report describes the analysis and modeling approaches used in the evaluation for criticality-control applications of the neutron-absorbing structural-amorphous metal (SAM) coatings. The applications of boron-containing high-performance corrosion-resistant material (HPCRM)--amorphous metal as the neutron-absorbing coatings to the metallic support structure can enhance criticality safety controls for spent nuclear fuel in baskets inside storage containers, transportation casks, and disposal containers. The use of these advanced iron-based, corrosion-resistant materials to prevent nuclear criticality in transportation, aging, and disposal containers would be extremely beneficial to the nuclear waste management programs.

Choi, J

2007-01-12T23:59:59.000Z

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


321

NNSA and Defense Nuclear Facilities Safety Board certifications free up $47  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas Conchas recoveryLaboratory | National Nuclear Security Administration Weaponsmillion in

322

Assessment of inservice conditions of safety-related nuclear plant structures  

SciTech Connect (OSTI)

The report is a compilation from a number of sources of information related to the condition Of structures and civil engineering features at operating nuclear power plants in the United States. The most significant information came from the hands-on inspection of the six old plants (licensed prior to 1977) performed by the staff of the Civil Engineering and Geosciences Branch (ECGB) in the Division of Engineering of the Office of Nuclear Reactor Regulation. For the containment structures, most of the information related to the degraded conditions came from the licensees as part of the Licensing Event Report System (10 CFR 50.73), or as part of the requirement under limiting condition of operation of the plant-specific Technical Specifications. Most of the information related to the degradation of other Structures and civil engineering features was extracted from the industry survey, the reported incidents, and the plant visits. The report discusses the condition of the structures and civil engineering features at operating nuclear power plants and provides information that would help detect, alleviate, and correct the degraded conditions of the structures and civil engineering features.

Ashar, H.; Bagchi, G.

1995-06-01T23:59:59.000Z

323

Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies  

SciTech Connect (OSTI)

Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study.

Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

1980-05-01T23:59:59.000Z

324

Nuclear Safety Research and Development Program Proposal Submittal Instructions for Fiscal Year 2015  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAXBalanced ScorecardReactor Technology Subcommittee of NEAC MujidDepartmentof Nuclear

325

Nuclear Safety Research and Development (NSR&D) Program | Department of  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission, Office

326

RESEARCH SAFETY RADIATION SAFETY  

E-Print Network [OSTI]

RESEARCH SAFETY RADIATION SAFETY ENVIRONMENTAL PROGRAMS HAZARDOUS MATERIALS CONTROLLED SUBSTANCES INTEGRATED WASTE MANAGEMENT LABORATORY SAFETY AUDITS & COMPLIANCE BIOSAFETY and ENVIRONMENTAL HEALTH EMERGENCY MANAGEMENT and MISSION CONTINUITY FIRE PREVENTION and LIFE SAFETY GENERAL SAFETY TRAINING

327

Technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1  

SciTech Connect (OSTI)

This report documents the technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1. The tests were to verify that faults on the non-Class 1E circuits would not propagate to the Class 1E circuits and degrade them below acceptable levels. The tests conducted demonstrated that the safety features actuation system did not degrade below acceptable levels nor was the system's ability to perform its protective functions affected.

Selan, J.C.

1981-07-01T23:59:59.000Z

328

Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2  

SciTech Connect (OSTI)

This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately.

Weiss, A. J. [comp.

1988-02-01T23:59:59.000Z

329

Report on the American Nuclear Society International Topical Meeting: {open_quotes}The safety, status, and future of non-commercial reactors and irradiation Facilities{close_quotes}  

SciTech Connect (OSTI)

The American Nuclear Society`s International Topical Meeting, The Safety, Status, and Future of Non-Commercial Reactors and Irradiation Facilities, also known as SAFOR 90, was held in Boise, Idaho, September 30 to October 4, 1990. In 19 half-day sessions, 102 papers were presented which covered operating research reactors, production reactors, the use of reactors for training and research, probabilistic risk assessments applied to research reactors, plans for new facilities, and new fuels and reactor types. A special session on space reactor safety was also presented. 11 refs., 1 tab.

Silver, E.G. [Oak Ridge National Laboratory, TN (United States)

1991-01-01T23:59:59.000Z

330

Analysis of Flow in Pilot Operated Safety and Relief Valve of Nuclear Reactor Coolant System  

SciTech Connect (OSTI)

When the POSRV equipped in a nuclear power plant opens in instant by a failure in coolant system of PWR, a moving shock wave generates, and propagates downstream of the valve, inducing a complicated unsteadiness. The moving shock wave may exert severe load to the structure. In this connection, a method of gradual opening of the valve is used to reduce the load acting on the wall at the downstream of the POSRV. In the present study, experiments and calculations are performed to investigate the detail unsteady flow at the various pipe units and the effect of valve opening time on the flow downstream of the valve. In calculation by using of air as working fluid, 2-dimensional, unsteady compressible Navier-Stokes equations are solved by finite volume method. It was found that when the incident shock wave passes through the pipe unit, it may experience diffraction, reflection and interaction with a vortex. Furthermore, the geometry of the pipe unit affects the reflection type of shock wave and changes the load acting on the wall of pipe unit. It was also turned out that the maximum force acting on the wall of the pipe unit becomes in order of T-junction, 108 deg. elbow and branch in magnitude, respectively. And, the results obtained that show that the rapid pressure rise due to the moving shock wave by instant POSRV valve opening is attenuated by employing the gradual opening. (authors)

Kwon, Soon-Bum; Lee, Dong-Won [Department of Mechanical Engineering, Kyungpook National University, 1370, Sankyuk-dong, Daegu 702-701 (Korea, Republic of); Kim, In-Goo; Ahn, Hyung-Joon; Kim, Hho-Jung [Korea Institute of Nuclear Safety, 19, Kusungdong, Yousungku, Daejon 305-338 (Korea, Republic of)

2004-07-01T23:59:59.000Z

331

Numerical solutions of the aerosol general dynamic equation for nuclear reactor safety studies  

SciTech Connect (OSTI)

Methods and approximations inherent in modeling of aerosol dynamics and evolution for nuclear reactor source term estimation have been investigated. Several aerosol evolution problems are considered to assess numerical methods of solving the aerosol dynamic equation. A new condensational growth model is constructed by generalizing Mason's formula to arbitrary particle sizes, and arbitrary accommodation of the condensing vapor and background gas at particle surface. Analytical solution is developed for the aerosol growth equation employing the new condensation model. The space-dependent aerosol dynamic equation is solved to assess implications of spatial homogenization of aerosol distributions. The results of our findings are as follows. The sectional method solving the aerosol dynamic equation is quite efficient in modeling of coagulation problems, but should be improved for simulation of strong condensation problems. The J-space transform method is accurate in modeling of condensation problems, but is very slow. For the situation considered, the new condensation model predicts slower aerosol growth than the corresponding isothermal model as well as Mason's model, the effect of partial accommodation is considerable on the particle evolution, and the effect of the energy accommodation coefficient is more pronounced than that of the mass accommodation coefficient. For the initial conditions considered, the space-dependent aerosol dynamics leads to results that are substantially different from those based on the spatially homogeneous aerosol dynamic equation.

Park, J.W.

1988-01-01T23:59:59.000Z

332

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

in U. S. Conunercial Nuclear Power Plants", Report WASH-Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"

Nero, A.V.

2010-01-01T23:59:59.000Z

333

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.  

SciTech Connect (OSTI)

This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

2011-02-01T23:59:59.000Z

334

A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants  

SciTech Connect (OSTI)

The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

1997-08-01T23:59:59.000Z

335

Operating experience feedback report-reliability of safety-related steam turbine-driven standby pumps used in US commerical nuclear power plants  

SciTech Connect (OSTI)

Pump failure experience is collected by two primary means: (1) Licensee Event Reports, and (2) Nuclear Plant Reliability Data System failure reports. Certain safety-related turbine-driven standby pumps were identified by these data systems as experiencing significant ongoing repetitive failures of their turbine drivers, resulting in low reliability of the pump units. The root causes of identified failures were determined, and actions to preclude these repetitive failures were identified. 5 refs., 1 tab.

Boardman, J.R. [Nuclear Regulatory Commission, Washington, DC (United States)

1995-01-01T23:59:59.000Z

336

Senior Technical Safety Manager Qualification Program Self-Assessment...  

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

Senior Technical Safety Manager Qualification Program Self-Assessment - Chief of Nuclear Safety Senior Technical Safety Manager Qualification Program Self-Assessment - Chief of...

337

Proceedings of the Advisory Committee on Reactor Safeguards Safety...  

Broader source: Energy.gov (indexed) [DOE]

Early Signs of Deteriorating Safety Performance (Thomas E. Murley, Safety Consultant) .33 2.1.4 The Role of the Nuclear Regulator in Promoting and Evaluating Safety Culture...

338

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA Energy and Environment

Nero, A.V.

2010-01-01T23:59:59.000Z

339

Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).  

SciTech Connect (OSTI)

This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

2010-09-01T23:59:59.000Z

340

Order Module--DOE O 452.1D, NUCLEAR EXPLOSIVE AND WEAPON SURETY PROGRAM, DOE O 452.2D, NUCLEAR EXPLOSIVE SAFETY  

Broader source: Energy.gov [DOE]

"To prevent accidents and inadvertent or unauthorized use of U.S. nuclear weapons and nuclear explosives. In conjunction with the Department of Defense (DoD), to protect the public health and...

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


341

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Report LBL-5287. "Power Plant Reliability-Availability andConunercial Nuclear Power Plants", Report WASH-1400 (NUREG-Standards for Nuclear Power Plants," by A.V. Nero and Y.C.

Nero, A.V.

2010-01-01T23:59:59.000Z

342

Nuclear Explosives Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This draft has been scheduled for final review before the Directives Review Board on 10/16/14. All major comments and concerns should be provided to your DRB representative, following your organization process. If you do not know who your representative is, please see the list of DRB members at https://www.directives.doe.gov/beta/references/directives-review-board. If your office is represented by Ingrid Kolb, Director, Office of Management, please submit your major concerns and comments to the DRB Liaison, Camille Beben (Camille.Beben@hq.doe.gov; 202-586-1014). All major comments and concerns should be submitted by COB 10/16/2014.

2014-10-16T23:59:59.000Z

343

Nuclear Safety Analysis Reports  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Cancels DOE O 5481.1B; paragraphs 7b(3), 7e(3) & 8c of DOE O 5480.6; and 51, 7b(3), 7b(4), 7e(3), 8a & 8h of DOE O 5480.5.

1992-04-30T23:59:59.000Z

344

Nuclear Safety Regulatory Framework  

Broader source: Energy.gov (indexed) [DOE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Careerlumens_placard-green.epsEnergy Second Quarter4, 2014Reactor TechnologyAugust 18,

345

Nuclear health and safety  

SciTech Connect (OSTI)

Since 1943 the Washington State Department of Labor and Industries has had a contract with the Department of Energy (DOE) or its predecessor to administer a self-insured workers' compensation/pension program for contractor employees at DOE's Hanford Site near Richland, Washington. This paper review stems from concerns that the contract's implementation could have prevented Hanford employees from filing workers' compensation claims for radiation-related injuries or occupational diseases resulting from their employment at the Hanford Site. GAO found that the procedures since the late 1950s for filing claims contain sufficient checks and balances to ensure they cannot be blocked by DOE. However, this assurance is lacking for claims initiated between 1943, when Hanford was founded, and the late 1950s. Claim-filed procedures in effect at that time required claims to the state before the procedural change, nor were DOE, state officials, or employee union representatives aware of any Hanford employee being denied the right to file a workers' compensation claim.

Not Available

1991-09-01T23:59:59.000Z

346

WIPP Documents - Nuclear Safety  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation Desert SouthwestTechnologies | Blandine Jerome VelenciaNewsNewsFederal RegulationsLabel

347

Radioisotope Power System Delivery, Ground Support and Nuclear Safety Implementation: Use of the Multi-Mission Radioisotope Thermoelectric Generator for the NASA's Mars Science Laboratory  

SciTech Connect (OSTI)

Radioisotope power systems have been used for over 50 years to enable missions in remote or hostile environments. They are a convenient means of supplying a few milliwatts up to a few hundred watts of useable, long-term electrical power. With regard to use of a radioisotope power system, the transportation, ground support and implementation of nuclear safety protocols in the field is a complex process that requires clear identification of needed technical and regulatory requirements. The appropriate care must be taken to provide high quality treatment of the item to be moved so it arrives in a condition to fulfill its missions in space. Similarly it must be transported and managed in a manner compliant with requirements for shipment and handling of special nuclear material. This presentation describes transportation, ground support operations and implementation of nuclear safety and security protocols for a radioisotope power system using recent experience involving the Multi-Mission Radioisotope Thermoelectric Generator for National Aeronautics and Space Administration’s Mars Science Laboratory, which launched in November of 2011.

S.G. Johnson; K.L. Lively; C.C. Dwight

2014-07-01T23:59:59.000Z

348

Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants  

SciTech Connect (OSTI)

Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail.

Staunton, R.H.; Cox, D.F. [Oak Ridge National Lab., TN (United States)

1995-03-01T23:59:59.000Z

349

National Nuclear Security Administration  

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

the safety, security and effectiveness of the nuclear deterrent without underground nuclear testing, consistent with the principles of the Stockpile Management Program...

350

Technical note: Erroneous data in {open_quotes}Nuclear Safety Guide, TID-7016, Revision 2,{close_quotes} NUREG/CR-0095, ORNL/NUREG/CSD-6 (1978)  

SciTech Connect (OSTI)

Nuclear Safety Guide, TID-7016 Rev 2 was issued as NUREG/CR-0095 in 1978. Table 2.8 of this report has been found to contain errors. The table was designed to indicate configurations with effective multipliction factors of 0.95. Because of an error in configuration descriptions, some of the configurations have multiplication factors as high as 1.09. A corrected table is available from the undersigned, and Revision 3 of the report is being prepared. Norman L. Pruvost, LANL HSE-6, P. O. Box 1663, M/S-F691, Los Alamos, NM 87545.

Whitesides, G.E. [Oak Ridge National Lab., TN (United States)

1991-04-01T23:59:59.000Z

351

Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

DOE-STD-1104 contains the Department's method and criteria for reviewing and approving nuclear facility's documented safety analysis (DSA). This review and approval formally document the basis for DOE, concluding that a facility can be operated safely in a manner that adequately protects workers, the public, and the environment. Therefore, it is appropriate to formally require implementation of the review methodology and criteria contained in DOE-STD-1104.

2013-06-21T23:59:59.000Z

352

Nuclear criticality safety evaluation of {sup 233}U storage configurations using ENDF/B-V cross sections  

SciTech Connect (OSTI)

Uranium-233 is currently stored in various chemical and physical forms in the Radiochernical Processing Plant (building 3019) and the Molten Salt Reactor (MSR) Facility (building 7503) at Oak Ridge National Laboratory (ORNL). Criticality safety is an important concern that must be addressed in the storage of this fissile material for both normal and credible abnormal conditions. The purpose of the current work is to perform a criticality safety evaluation of the {sup 233}U inventory at ORNL using KENO V.a with ENDF/B-V cross sections.

Dunn, M.E.; Basoglu, B.; Bentley, C.L.; Goluoglu, S.; Haught, C.; Plaster, M.J.; Wilkinson, A.D.; Yamamoto, T.; Dodds, H.L. [Univ. of Tennessee, Knoxville, TN (United States)

1994-12-31T23:59:59.000Z

353

Pantex Plant final safety analysis report, Zone 4 magazines. Staging or interim storage for nuclear weapons and components: Issue D  

SciTech Connect (OSTI)

This Safety Analysis Report (SAR) contains a detailed description and evaluation of the significant environmental, safety, and health (ES&H) issues associated with the operations of the Pantex Plant modified-Richmond and steel arch construction (SAC) magazines in Zone 4. It provides (1) an overall description of the magazines, the Pantex Plant, and its surroundings; (2) a systematic evaluations of the hazards that could occur as a result of the operations performed in these magazines; (3) descriptions and analyses of the adequacy of the measures taken to eliminate, control, or mitigate the identified hazards; and (4) analyses of potential accidents and their associated risks.

Not Available

1993-04-01T23:59:59.000Z

354

Review guidelines for software languages for use in nuclear power plant safety systems: Final report. Revision 1  

SciTech Connect (OSTI)

Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada83 and Ada95; C and C++; International Electrochemical Commission (IEC) Standard 1131-3 Ladder Logic, Sequential Function Charts, Structured Text, and Function Block Diagrams; Pascal; and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.

Hecht, M.; Decker, D.; Graff, S.; Green, W.; Lin, D.; Dinsmore, G.; Koch, S. [SoHaR, Inc., Beverly Hills, CA (United States)

1997-10-01T23:59:59.000Z

355

In-Situ Radiological Surveys to Address Nuclear Criticality Safety Requirements During Remediation Activities at the Shallow Land Disposal Area, Armstrong County, Pennsylvania - 12268  

SciTech Connect (OSTI)

Cabrera Services Inc. (CABRERA) is the remedial contractor for the Shallow Land Disposal Area (SLDA) Site in Armstrong County Pennsylvania, a United States (US) Army Corps of Engineers - Buffalo District (USACE) contract. The remediation is being completed under the USACE's Formerly Utilized Sites Remedial Action Program (FUSRAP) which was established to identify, investigate, and clean up or control sites previously used by the Atomic Energy Commission (AEC) and its predecessor, the Manhattan Engineer District (MED). As part of the management of the FUSRAP, the USACE is overseeing investigation and remediation of radiological contamination at the SLDA Site in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), 42 US Code (USC), Section 9601 et. seq, as amended and, the National Oil and Hazardous Substance Pollution Contingency Plan (NCP), Title 40 of the Code of Federal Regulations (CFR) Section 300.430(f) (2). The objective of this project is to clean up radioactive waste at SLDA. The radioactive waste contains special nuclear material (SNM), primarily U-235, in 10 burial trenches, Cabrera duties include processing, packaging and transporting the waste to an offsite disposal facility in accordance with the selected remedial alternative as defined in the Final Record of Decision (USACE, 2007). Of particular importance during the remediation is the need to address nuclear criticality safety (NCS) controls for the safe exhumation and management of waste containing fissile materials. The partnership between Cabrera Services, Inc. and Measutronics Corporation led to the development of a valuable survey tool and operating procedure that are essential components of the SLDA Criticality Safety and Material Control and Accountability programs. Using proven existing technologies in the design and manufacture of the Mobile Survey Cart, the continued deployment of the Cart will allow for an efficient and reliable methodology to allow for the safe exhumation of the Special Nuclear Material in existing SLDA trenches. (authors)

Norris, Phillip; Mihalo, Mark; Eberlin, John; Lambert, Mike [Cabrera Services (United States); Matthews, Brian [Nuclear Safety Associates (United States)

2012-07-01T23:59:59.000Z

356

A nuclear criticality safety assessment of the loss of moderation control in 2 1/2 and 10-ton cylinders containing enriched UF{sub 6}  

SciTech Connect (OSTI)

Moderation control for maintaining nuclear criticality safety in 2 {1/2}-ton, 10-ton, and 14-ton cylinders containing enriched uranium hexafluoride (UF{sub 6}) has been used safely within the nuclear industry for over thirty years, and is dependent on cylinder integrity and containment. This assessment evaluates the loss of moderation control by the breaching of containment and entry of water into the cylinders. The first objective of this study was to estimate the required amounts of water entering these large UF{sub 6} cylinders to react with, and to moderate the uranium compounds sufficiently to cause criticality. Hypothetical accident situations were modeled as a uranyl fluoride (UO{sub 2}F{sub 2}) slab above a UF{sub 6} hemicylinder, and a UO{sub 2}F{sub 2} sphere centered within a UF{sub 6} hemicylinder. These situations were investigated by computational analyses utilizing the KENO V.a Monte Carlo Computer Code. The results were used to estimate both the masses of water required for criticality, and the limiting masses of water that could be considered safe. The second objective of the assessment was to calculate the time available for emergency control actions before a criticality would occur, i.e., a {open_quotes}safetime{close_quotes}, for various sources of water and different size openings in a breached cylinder. In the situations considered, except the case for a fire hose, the safetime appears adequate for emergency control actions. The assessment shows that current practices for handling moderation controlled cylinders of low enriched UF{sub 6}, along with the continuation of established personnel training programs, ensure nuclear criticality safety for routine and emergency operations.

Newvahner, R.L. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States)

1991-12-31T23:59:59.000Z

357

Summary of the contractor information exchange meeting for improving the safety of Soviet-Designed Nuclear Power Plants, February 19, 1997  

SciTech Connect (OSTI)

This report summarizes a meeting held on February 19, 1997, in Washington, D.C. The meeting was held primarily to exchange information among the contractors involved in the U.S. Department of Energy`s efforts to improve the safety of Soviet-designed nuclear power plants. Previous meetings have been held on December 5-6, 1995, and May 22, 1996. The meetings are sponsored by the U.S. Department of Energy and coordinated by the Pacific Northwest National Laboratory. The U.S. Department of Energy works with countries to increase the level of safety at 63 Soviet-designed nuclear reactors operating in Armenia, Bulgaria, the Czech Republic, Hungary, Lithuania, Russia, Slovakia, and Ukraine. The work is implemented largely by commercial companies and individuals who provide technologies and services to the countries with Soviet-designed nuclear power plants. Attending the meeting were 71 representatives of commercial contractors, the U.S. Department of Energy, the U.S. Department of State, national laboratories, and other federal agencies. The presentations and discussions that occurred during the exchange are summarized in this report. While this report captures the general presentation and discussion points covered at the meeting, it is not a verbatim, inclusive record. To make the report useful, information presented at the meeting has been expanded to clarify issues, respond to attendees` requests, or place discussion points in a broader programmatic context. Appendixes A through F contain the meeting agenda, list of attendees, copies of presentation visuals and handouts, the Strategy Document discussed at the meeting, and a summary of attendees` post-meeting evaluation comments. As with past information exchanges, the participants found this meeting valuable and useful. In response to the participant`s requests, a fourth information exchange will be held later in 1997.

NONE

1997-04-01T23:59:59.000Z

358

Worker Safety and Health Program for DOE (Including the National Nuclear Security Administration) Federal and Contractor Employees  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This document was developed to assist the Department of Energy (DOE or the Department) Federal and contractor employees in effectively developing, managing, and implementing a worker safety and health program. Cancels DOE G 440.1-1A and DOE G 440.1-8. Adm Chg 1, dated 3-22-13.

2011-10-20T23:59:59.000Z

359

Worker Safety and Health Program for DOE (Including the National Nuclear Security Administration) Federal and Contractor Employees  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This document was developed to assist the Department of Energy (DOE or the Department) Federal and contractor employees in effectively developing, managing, and implementing a worker safety and health program. Cancels DOE G 440.1-1A and DOE G 440.1-8. Adm Chg 1, dated 3-22-13, cancels DOE G 440.1-1B.

2011-10-20T23:59:59.000Z

360

Safety research programs sponsored by Office of Nuclear Regulatory Research: Progress report, October 1--December 31, 1987  

SciTech Connect (OSTI)

The Advanced Reactor Review Safety Research Program has continued under NRC Research funding since October 1986. In April 1987, as a result of an NRC reorganization, the Accident Analysis and Safety Review of Liquid Metal and High Temperature Gas Reactors (LMRs and HTGRs) Program was transferred from Regulation to Research. Both programs are discussed here under the title of Accident Analysis and Safety Review of Liquid Metal and High Temperature Gas Reactors (LMRs and HTGRs). The combined programs are focused to help NRC accomplish the review of advanced reactor (LMR and HTGR) designs over the next year. Technical assistance in the following areas is provided: (1) review, adaptation, and implementation of analytical tools and models for application to the design submittals, (2) independent analysis of specific accidents and plant conditions and characteristics for the designs, (3) developing and evaluating appropriate source terms, (4) reviewing DOE reports on safety issues, (5) review of those aspects of DOE's base technical program related to the areas described in (1) through (4), and (6) assistance in assessing PRAs submitted for review.

Weiss, A.J. (comp.); Bari, R.A.; Boccio, J.L.; Fitzpatrick, R.; Ginsberg, T.; Greene, G.A.; Guppy, J.G.; Hall, R.E.; Higgins, J.C.; Khatib-Rahbar, M.

1988-08-01T23:59:59.000Z

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


361

Risk-informed public safety policy for seismic events in the vicinity of a nuclear power plant  

E-Print Network [OSTI]

Nuclear Power Plants (NPPs) are potentially vulnerable to accidents, which can either be internally or externally initiated. External events include natural events like tornadoes, hurricanes, and earthquakes. The purpose ...

Afolayan Jejeloye, Olubukola

2002-01-01T23:59:59.000Z

362

Technical Basis for U. S. Department of Energy Nuclear Safety Policy, DOE Policy 420.1, 7/11  

Broader source: Energy.gov [DOE]

It is the policy of the Department of Energy (DOE) to design, construct, operate anddecommission its nuclear facilities in a manner that ensures adequate protection ofworkers, the public, and the...

363

Safety First Safety Last Safety Always Safety Shoes  

E-Print Network [OSTI]

Safety First Safety Last Safety Always Safety Shoes and Boots Safety Tip #21 Don't let your day guards) can be used in conjunction with standard safety shoes. Safety boots Safety boots come in many varieties, and which you will use will depend on the specific hazards you face. Boots offer more protection

Minnesota, University of

364

Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 15  

SciTech Connect (OSTI)

This report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), Supplement No. 8 (January 1992), Supplement No. 9 (June 1992), Supplement No. 10 (October 1992), Supplement No. 11 (April 1993), Supplement No. 12 (October 1993), Supplement No. 13 (April 1994), and Supplement No. 14 (December 1994) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER.

Tam, P.S.

1995-06-01T23:59:59.000Z

365

Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Numbers 50-390 and 50-391). Supplement Number 13  

SciTech Connect (OSTI)

This report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), Supplement No. 8 (January 1992), Supplement No. 9 (June 1992), Supplement No. 10 (October 1992), Supplement No. 11 (April 1993), and Supplement No. 12 (October 1993), issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER. These issues relate to: Design criteria -- structures, components, equipment, and systems; Reactor; Instrumentation and controls; Electrical power systems; Auxiliary systems; Conduct of operations; Accident analysis; and Quality assurance.

Not Available

1994-04-01T23:59:59.000Z

366

Thermal reactor safety  

SciTech Connect (OSTI)

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

Not Available

1980-06-01T23:59:59.000Z

367

Safety of Accelerator Facilities  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establish accelerator-specific safety requirements which, when supplemented by other applicable safety and health requirements, will serve to prevent injuries and illnesses associated with Department of Energy (DOE) or National Nuclear Security Administration (NNSA) accelerator operations. Cancels DOE O 420.2A. Certified 5-13-08. Canceled by DOE O 420.2C.

2004-07-23T23:59:59.000Z

368

Safety of Accelerator Facilities  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establish accelerator-specific safety requirements which, when supplemented by other applicable safety and health requirements, will serve to prevent injuries and illnesses associated with Department of Energy (DOE) or National Nuclear Security Administration (NNSA) accelerator operations. Cancels DOE O 420.2. Canceled by DOE O 420.2B.

2001-01-08T23:59:59.000Z

369

Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 1, Plenary sessions, reactor licensing topics, NUREG-1150, risk analysis/PRA applications, innovative concepts for increased safety of advanced power reactors, severe accident modeling and analysis  

SciTech Connect (OSTI)

This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 1, discusses the following: plenary sessions; reactor licensing; NUREG-1150; risk analysis; innovative concepts for increased safety of advanced power reactors; and severe accident modeling and analysis. Thirty-two reports have been cataloged separately.

Weiss, A.J. (comp.)

1988-02-01T23:59:59.000Z

370

A Proposed Cost-Benefit Analysis Approach for Evaluating DOE Nuclear Facility Design Options  

Broader source: Energy.gov [DOE]

Presenter: Dr. Kamiar Jamali, Senior Technical Advisor to the Chief of Defense Nuclear Safety, National Nuclear Security Administration, Office of Nuclear Safety NA-SH

371

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

for sabotage of nuclear plants or the potential forAccidents in Large Nuclear Plants" (AEC Report, WASH-/40),YEAR FROM POTENTIAL NUCLEAR PLANT ACCIDENTS(a) Consequence

Nero, A.V.

2010-01-01T23:59:59.000Z

372

Sandia National Laboratories: Nuclear Energy Publications  

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

Publications Nuclear Energy Publications Nuclear Energy Safety Fact Sheets Assuring Safe Transportation of Nuclear and Hazardous Materials Human Reliability Assessment (HRA)...

373

RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

EMERGENCY PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSINGEmergency Planning for Nuclear Power Plants Determination ofproposed nuclear power plants . . . . . . . . . • . . . .

Yen, W.W.S.

2010-01-01T23:59:59.000Z

374

Radiation Safety  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared at 278, 298, and 323 RS-PO-0001-001.docW. J:.EnergySafety Home

375

Safety First Safety Last Safety Always General site safety  

E-Print Network [OSTI]

Safety First Safety Last Safety Always General site safety During the course of construction barrier at least 5 feet (1.5m) high having a fire-resistance rating of at least one half hour. Site Safety and Clean-up Safety Tip #20 Safety has no quitting time. All contractors should clean up their debris, trash

Minnesota, University of

376

Safety First Safety Last Safety Always Safety Tip #22  

E-Print Network [OSTI]

Safety First Safety Last Safety Always Safety Tip #22 Mowing Operations Mowing unsafely just doesn for out-of-control vehicles. Wear hearing protection and a safety vest. Wear a hard hat and safety goggles of this safety tip sheet. Please refrain from reading the information verbatim--paraphrase it instead

Minnesota, University of

377

CRAD, Criteria and Guidelines For the Assessment of Safety System...  

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

CRAD, Criteria and Guidelines For the Assessment of Safety System Software and Firmware at Defense Nuclear Facilities CRAD, Criteria and Guidelines For the Assessment of Safety...

378

Key Practical Issues in Strengthening Safety Culture, INSAG-15  

Broader source: Energy.gov [DOE]

Key Pratical Issues in Strengthening Safety Culture, INSAG-15. A report by the International Nuclear Safety Advisory Gorup, International Atomic Energy Agency, Vienna, 2002.

379

Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980  

SciTech Connect (OSTI)

This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).

McCormack, K.E.; Gallaher, R.B.

1982-03-01T23:59:59.000Z

380

FLUOR HANFORD SAFETY MANAGEMENT PROGRAMS  

SciTech Connect (OSTI)

This document summarizes safety management programs used within the scope of the ''Project Hanford Management Contract''. The document has been developed to meet the format and content requirements of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses''. This document provides summary descriptions of Fluor Hanford safety management programs, which Fluor Hanford nuclear facilities may reference and incorporate into their safety basis when producing facility- or activity-specific documented safety analyses (DSA). Facility- or activity-specific DSAs will identify any variances to the safety management programs described in this document and any specific attributes of these safety management programs that are important for controlling potentially hazardous conditions. In addition, facility- or activity-specific DSAs may identify unique additions to the safety management programs that are needed to control potentially hazardous conditions.

GARVIN, L J; JENSEN, M A

2004-04-13T23:59:59.000Z

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


381

Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1979  

SciTech Connect (OSTI)

This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1979. The 1345 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Keyword and permuted-title indexes to facilitate location of individual abstracts are provided in full size following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of the failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Some of the more interesting events that occurred during the year are reviewed in detail. 32 refs.

Scott, R.L.; Gallaher, R.B.

1981-06-01T23:59:59.000Z

382

The (safety-related) heat exchangers aging management guideline for commercial nuclear power plants, and developments since 1994  

SciTech Connect (OSTI)

The US Department of Energy (DOE), in cooperation with the Electric Power Research Institute (EPRI) and US nuclear power plant utilities, is preparing a series of aging management guidelines (AMGs) for commodity types of components (e.g., heat exchangers, electrical cable and terminations, pumps). Commodities are included in this series based on their importance to continued nuclear plant operation and license renewal. The AMGs contain a detailed summary of operating history, stressors, aging mechanisms, and various types of maintenance and surveillance practices that can be combined to create an effective aging management program. Each AMG is intended for use by the systems engineers and plant maintenance staff (i.e., an AMG is intended to be a hands-on technical document rather than a licensing document). The heat exchangers AMG, published in June 1994, includes the following information of interest to nondestructive examination (NDE) personnel: aging mechanisms determined to be non-significant for all applications; aging mechanisms determined to be significant for some applications; effective conventional programs for managing aging; and effective unconventional programs for managing aging. Since the AMG on heat exchangers was published four years ago, a brief review has been conducted to identify emerging regulatory issues, if any. The results of this review and lessons learned from the collective set of AMGs are presented.

Clauss, J.M.

1998-08-01T23:59:59.000Z

383

Final safety analysis report for the Galileo mission: Volume 3 (Book 2), Nuclear risk analysis document: Appendices: Revision 1  

SciTech Connect (OSTI)

It is the purpose of the NRAD to provide an analysis of the range of potential consequences of accidents which have been identified that are associated with the launching and deployment of the Galileo mission spacecraft. The specific consequences analyzed are those associated with the possible release of radioactive material (fuel) of the Radioisotope Thermoelectric Generators (RTGs). They are in terms of radiation doses to people and areas of deposition of radioactive material. These consequence analyses can be used in several ways. One way is to identify the potential range of consequences which might have to be dealt with if there were to be an accident with a release of fuel, so as to assure that, given such an accident, the health and safety of the public will be reasonably protected. Another use of the information, in conjunction with accident and release probabilities, is to estimate the risks associated with the mission. That is, most space launches occur without incident. Given an accident, the most probable result relative to the RTGs is complete containment of the radioactive material. Only a small fraction of accidents might result in a release of fuel and subsequent radiological consequences. The combination of probability with consequence is risk, which can be compared to other human and societal risks to assure that no undue risks are implied by undertaking the mission. Book 2 contains eight appendices.

Not Available

1989-01-25T23:59:59.000Z

384

UNBC SAFETY CHECKLIST SAFETY CHECKLIST  

E-Print Network [OSTI]

1 UNBC SAFETY CHECKLIST SAFETY CHECKLIST INSTRUCTIONS PAGE Please use the following table below needs, contact the Risk & Safety Department at 250-960- (5530) for further instructions. This safety. The safety checklist also helps you to establish due diligence under Federal and Provincial safety laws

Northern British Columbia, University of

385

Toolbox Safety Talk Ladder Safety  

E-Print Network [OSTI]

Toolbox Safety Talk Ladder Safety Environmental Health & Safety Facilities Safety & Health Section Health & Safety for recordkeeping. Slips, trips, and falls constitute the majority of general industry elevated work tasks. Like any tool, ladders must be used properly to ensure employee safety. GENERAL

Pawlowski, Wojtek

386

Nuclear reactor engineering  

SciTech Connect (OSTI)

Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

Glasstone, S.; Sesonske, A.

1981-01-01T23:59:59.000Z

387

CRAD, Facility Safety- Unreviewed Safety Question Requirements  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Unreviewed Safety Question (USQ) process.

388

CRAD, Facility Safety- Technical Safety Requirements  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Technical Safety Requirments (TSA).

389

RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

S. Commercial Nuclear Power Plants. WASH-1400. October 1975.Content of for Nuclear Power Plants. Regulatory Guide 1.101.PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSING PROCESS

Yen, W.W.S.

2010-01-01T23:59:59.000Z

390

RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

from the Rancho Seco nuclear plant was simulated, A total ofdistributions around the nuclear plant sites based on thegrowth surrounding nuclear plants after the issuance of the

Yen, W.W.S.

2010-01-01T23:59:59.000Z

391

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

surrounding a nuclear plant, and they are stronglylocation for a nuclear plant, but it is the measures thatand consequences of nuclear plant accidents and would match

Nero, jA.V.

2010-01-01T23:59:59.000Z

392

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

IMPACTS OF FOSSIL-FUEL NUCLEAR, GEOTHERMAL, AND ELECTRIC GENERATION IN CALIFORNIA Energy and Environment

Nero, A.V.

2010-01-01T23:59:59.000Z

393

Exploration of High-dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization: A User's Guide to TopoXG*  

SciTech Connect (OSTI)

Large-scale simulation datasets can be modeled as high-dimensional scalar functions defined over a discrete sample of the domain. The goals of our proposed research are two-fold. First, we would like to provide structural analysis of a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. TopoXG is a software package that is designed to address these goals. The unique contribution of TopoXG lies in exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to TopoXG, by highlighting its analysis and visualization capabilities, and giving several use cases involving datasets from nuclear reactor safety simulations.

Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Diego Mandelli

2012-10-01T23:59:59.000Z

394

Safety First Safety Last Safety Always Requirements for employers  

E-Print Network [OSTI]

Safety First Safety Last Safety Always Requirements for employers · Fallprotectionsandproperuseofrelated-safety equipmentsuchaslifelines,harness · Properuseofdangeroustools,thenecessaryprecautionstotake,andtheuseof theprotectiveandemergencyequipmentrequired. Safety Training and Education Safety Tip #18 Get smart. Use safety from the start. All

Minnesota, University of

395

Criticality Safety Evaluation of a LLNL Training Assembly for Criticality Safety (TACS)  

SciTech Connect (OSTI)

Hands-on experimental training in the physical behavior of multiplying systems is one of ten key areas of training required for practitioners to become qualified in the discipline of criticality safety as identified in DOE-STD-1135-99, ''Guidance for Nuclear Criticality Safety Engineer Training and Qualification''. This document is a criticality safety evaluation of the training activities (or operations) associated with HS-3200, ''Laboratory Class for Criticality Safety''. These activities utilize the Training Assembly for Criticality Safety (TACS). The original intent of HS-3200 was to provide LLNL fissile material handlers with a practical hands-on experience as a supplement to the academic training they receive biennially in HS-3100, ''Fundamentals of Criticality Safety'', as required by ANSI/ANS-8.20-1991, ''Nuclear Criticality Safety Training''. HS-3200 is to be enhanced to also address the training needs of nuclear criticality safety professionals under the auspices of the NNSA Nuclear Criticality Safety Program.

Heinrichs, D P

2006-06-26T23:59:59.000Z

396

Office of Environment, Safety and Health Assessments Protocol...  

Office of Environmental Management (EM)

National Nuclear Security Administration CRAD, Radiological Controls - Idaho Accelerated Retrieval Project Phase II CRAD, Preliminary Documented Safety Analysis - July 25,...

397

Hanford safety analysis and risk assessment handbook (SARAH)  

SciTech Connect (OSTI)

The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 1,2, and 3 U.S. Department of Energy (DOE) nuclear facilities. SARAH describes currently acceptable methodology for development of a Documented Safety Analysis (DSA) and derivation of technical safety requirements (TSR) based on 10 CFR 830, ''Nuclear Safety Management,'' Subpart B, ''Safety Basis Requirements,'' and provides data to ensure consistency in approach.

GARVIN, L.J.

2003-01-20T23:59:59.000Z

398

Order Module--DOE-STD-1104-2009, REVIEW AND APPROVAL OF NUCLEAR...  

Broader source: Energy.gov (indexed) [DOE]

AND APPROVAL OF NUCLEAR FACILITY SAFETY BASIS AND SAFETY DESIGN BASIS DOCUMENTS Order Module--DOE-STD-1104-2009, REVIEW AND APPROVAL OF NUCLEAR FACILITY SAFETY BASIS AND SAFETY...

399

Seismic Safety Guide  

SciTech Connect (OSTI)

This guide provides managers with practical guidelines for administering a comprehensive earthquake safety program. The Guide is comprehensive with respect to earthquakes in that it covers the most important aspects of natural hazards, site planning, evaluation and rehabilitation of existing buildings, design of new facilities, operational safety, emergency planning, special considerations related to shielding blocks, non-structural elements, lifelines, fire protection and emergency facilities. Management of risk and liabilities is also covered. Nuclear facilities per se are not dealt with specifically. The principles covered also apply generally to nuclear facilities but the design and construction of such structures are subject to special regulations and legal controls.

Eagling, D.G. (ed.)

1983-09-01T23:59:59.000Z

400

UNBC SAFETY CHECKLIST SAFETY CHECKLIST  

E-Print Network [OSTI]

1 UNBC SAFETY CHECKLIST SAFETY CHECKLIST INSTRUCTIONS PAGE Please use the following table below needs, contact the Risk & Safety Department at 250-960- (5530) for further instructions. This safety to remain safe here at UNBC. The safety checklist also helps you to establish due diligence under Federal

Northern British Columbia, University of

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


401

China's Nuclear Industry After Fukushima  

E-Print Network [OSTI]

Brief 2013-9 January 2013 China’s Nuclear Industry Aftera significant impact on the future of China’s nuclear power.the importance of safety as China builds more nuclear power

YUAN, Jingdong

2013-01-01T23:59:59.000Z

402

Integrated Safety Management in QA Program Planning  

Broader source: Energy.gov [DOE]

Presenter: Sonya Barnette, Office of Quality Assurance Policy and Assistance, Office of Nuclear Safety, Quality Assurance and Environment Track 9-8

403

Use of InSpector{sup TM} 1 1000 Instrument with LaBr{sub 3} for Nuclear Criticality Safety (NCS) Applications at the Westinghouse Hematite Decommissioning Project (HDP) - 13132  

SciTech Connect (OSTI)

The Westinghouse Hematite Decommissioning Project (HDP) is a former nuclear fuel cycle facility that is currently undergoing decommissioning. One aspect of the decommissioning scope is remediation of buried nuclear waste in unlined burial pits. The current Nuclear Criticality Safety program relies on application of criticality controls based on radiological setpoints from a 2 x 2 Sodium Iodide (NaI) detector. Because of the nature of the material buried (Low Enriched Uranium (LEU), depleted uranium, thorium, and radium) and the stringent threshold for application of criticality controls based on waste management (0.1 g {sup 235}U/L), a better method for {sup 235}U identification and quantification has been developed. This paper outlines the early stages of a quick, in-field nuclear material assay and {sup 235}U mass estimation process currently being deployed at HDP. Nuclear material initially classified such that NCS controls are necessary can be demonstrated not to require such controls and dispositioned as desired by project operations. Using Monte Carlo techniques and a high resolution Lanthanum Bromide (LaBr) detector with portable Multi-Channel Analyzer (MCA), a bounding {sup 235}U mass is assigned to basic geometries of nuclear material as it is excavated. The deployment of these methods and techniques has saved large amounts of time and money in the nuclear material remediation process. (authors)

Pritchard, Megan [Nuclear Safety Associates, P.O. Box 471488, Charlotte, NC 28247 (United States)] [Nuclear Safety Associates, P.O. Box 471488, Charlotte, NC 28247 (United States); Guido, Joe [System One Services, 12 Federal St. Ste. 205, Pittsburgh, PA 15212 (United States)] [System One Services, 12 Federal St. Ste. 205, Pittsburgh, PA 15212 (United States)

2013-07-01T23:59:59.000Z

404

Toolbox Safety Talk Safety Data Sheets (SDS)  

E-Print Network [OSTI]

Toolbox Safety Talk Safety Data Sheets (SDS) Environmental Health & Safety Facilities Safety-in sheet to Environmental Health & Safety for recordkeeping. Chemical manufacturers are required to produce Safety Data Sheets (SDS) for all chemicals produced. "Safety Data Sheets", previously referred

Pawlowski, Wojtek

405

Nuclear Explosive and Weapon Surety Program - DOE Directives...  

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

1E, Nuclear Explosive and Weapon Surety Program by Angela Chambers Functional areas: Defense Nuclear Facility Safety and Health Requirement, Defense Programs, Nuclear Weapons...

406

Nuclear Explosive and Weapon Surety Program - DOE Directives...  

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

D, Nuclear Explosive and Weapon Surety Program by cdornburg Functional areas: Defense Nuclear Facility Safety and Health Requirement, Defense Programs, Nuclear Weapons Programs,...

407

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Nuclear Power Reactors PROTECTION AGAINST SABOTAGE Protection Against Industrial Sabotage I1C-4 Decominarion and Decommissioning

Nero, A.V.

2010-01-01T23:59:59.000Z

408

On the quantification of safety margins  

E-Print Network [OSTI]

The nuclear industry has relied on the concept of Defense in Depth (DID) and traditional safety margins to deal with the uncertainties associated with the design and operation of nuclear facilities. These concepts were ...

Pagani, Lorenzo P

2004-01-01T23:59:59.000Z

409

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) verification and validation plan. version 1.  

SciTech Connect (OSTI)

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. To meet this objective, NEAMS Waste IPSC M&S capabilities will be applied to challenging spatial domains, temporal domains, multiphysics couplings, and multiscale couplings. A strategic verification and validation (V&V) goal is to establish evidence-based metrics for the level of confidence in M&S codes and capabilities. Because it is economically impractical to apply the maximum V&V rigor to each and every M&S capability, M&S capabilities will be ranked for their impact on the performance assessments of various components of the repository systems. Those M&S capabilities with greater impact will require a greater level of confidence and a correspondingly greater investment in V&V. This report includes five major components: (1) a background summary of the NEAMS Waste IPSC to emphasize M&S challenges; (2) the conceptual foundation for verification, validation, and confidence assessment of NEAMS Waste IPSC M&S capabilities; (3) specifications for the planned verification, validation, and confidence-assessment practices; (4) specifications for the planned evidence information management system; and (5) a path forward for the incremental implementation of this V&V plan.

Bartlett, Roscoe Ainsworth; Arguello, Jose Guadalupe, Jr.; Urbina, Angel; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Knupp, Patrick Michael; Wang, Yifeng; Schultz, Peter Andrew; Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); McCornack, Marjorie Turner

2011-01-01T23:59:59.000Z

410

Correlation of nuclear criticality safety computer codes with plutonium benchmark experiments and derivation of subcritical limits. [MGBS, TGAN, KEFF, HRXN, GLASS, ANISN, SPBL, and KENO  

SciTech Connect (OSTI)

A compilation of benchmark critical experiments was made for essentially one-dimensional systems containing plutonium. The systems consist of spheres, series of experiments with cylinders and cuboids that permit extrapolation to infinite cylinders and slabs, and large cylinders for which separability of the neutron flux into a product of spatial components is a good approximation. Data from the experiments were placed in a form readily usable as computer code input. Aqueous solutions of Pu(NO/sub 3/)/sub 4/ are treated as solutions of PuO/sub 2/ in nitric acid. The apparent molal volume of PuO/sub 2/ as a function of plutonium concentration was derived from analyses of solution density data and was incorporated in the Savannah River Laboratory computer codes along with density tables for nitric acid. The biases of three methods of calculation were established by correlation with the benchmark experiments. The oldest method involves two-group diffusion theory and has been used extensively at the Savannah River Laboratory. The other two involve S/sub n/ transport theory with, in one method, Hansen-Roach cross sections and, in the other, cross sections derived from ENDF/B-IV. Subcritical limits were calculated by all three methods. Significant differences were found among the results and between the results and limits currently in the American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactor (ANSI N16.1), which were calculated by yet another method, despite the normalization of all four methods to the same experimental data. The differences were studied, and a set of subcritical limits was proposed to supplement and in some cases to replace those in the ANSI Standard, which is currently being reviewed.

Clark, H.K.

1981-10-01T23:59:59.000Z

411

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"Densities Surrounding Nuclear Power Plants," by A.V. Nero,

Nero, jA.V.

2010-01-01T23:59:59.000Z

412

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Standards for Nuclear Power Plants," by A.V. Nero and Y.C.Planning for Nuclear Power Plants in California," by W.W.S.Surrounding Nuclear Power Plants," by A.V. Nero, C.H.

Nero, jA.V.

2010-01-01T23:59:59.000Z

413

METHODOLOGIES FOR REVIEW OF THE HEALTH AND SAFETY ASPECTS OF PROPOSED NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL SITES AND FACILITIES. VOLUME 9 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

emergencies, Le. , accidents at nuclear facilities, there isas a result of nuclear accidents; these are the Protectiveassociated with a nuclear accident is of greater importance

Nero, A.V.

2010-01-01T23:59:59.000Z

414

Reactor safety method  

DOE Patents [OSTI]

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

415

Packaging and Transportation Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

To establish safety requirements for the proper packaging and transportation of Department of Energy (DOE)/National Nuclear Security Administration (NNSA) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1A. Canceled by DOE O 460.1C.

2003-04-04T23:59:59.000Z

416

Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors  

E-Print Network [OSTI]

decommissioning. For example, methodologies that are common practice for the safety analysis of nuclear

Scarlat, Raluca Olga

2012-01-01T23:59:59.000Z

417

Nuclear Reactions and Reactor Safety  

E-Print Network [OSTI]

. Snoeijer 85 On the drag of turbulent vortex rings L. Gan, J. R. Dawson & T. B. Nickels 106 Subcritical Linear water waves: the horizontal motion of a structure in the time domain P. McIver 313 Transmission

Onuchic, José

418

Nuclear Explosive Safety Study Process  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAXBalanced Scorecard Federal2EnergyDepartment511LawsVerificationResearchi

419

Nuclear Safety | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGYWomen OwnedofDepartment ofJaredOakscience-based, applied engineeringTVA WattsOffice of

420

Defense Nuclear Facility Safety Board  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation Proposed Newcatalyst phasesData Files Data Files 1B&W Y-12studies in ionand Minor8, 2014

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


422

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA Energy and Environment

Nero, jA.V.

2010-01-01T23:59:59.000Z

423

Implementation Guide for Use in Developing Technical Safety Requirements  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Guide provides elaboration for the content of TSRs. Section 10 CFR 830.205 of the Nuclear Safety Management rule, requires Department of Energy (DOE) contractors responsible for category 1, 2, and 3 DOE nuclear facilities to develop Technical Safety Requirements (TSRs). These TSRs identify the limitations to each DOE owned, contractor operated nuclear facility based on the documented safety analysis (DSA) and any additional safety requirements established for the facility. Does not cancel other directives.

2010-11-03T23:59:59.000Z

424

Implementation Guide for Use in Developing Technical Safety Requirements  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Guide provides elaboration for the content of TSRs. Section 10 CFR 830.205 of the Nuclear Safety Management rule, requires Department of Energy (DOE) contractors responsible for category 1, 2, and 3 DOE nuclear facilities to develop Technical Safety Requirements (TSRs). These TSRs identify the limitations to each DOE owned, contractor operated nuclear facility based on the documented safety analysis (DSA) and any additional safety requirements established for the facility. Cancels DOE G 423.1-1.

2010-11-03T23:59:59.000Z

425

Chemical Safety Vulnerability Working Group Report  

SciTech Connect (OSTI)

This report marks the culmination of a 4-month review conducted to identify chemical safety vulnerabilities existing at DOE facilities. This review is an integral part of DOE's efforts to raise its commitment to chemical safety to the same level as that for nuclear safety.

Not Available

1994-09-01T23:59:59.000Z

426

Facility Safety (9-23-10)--Withdrawn  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

Withdrawn, 5-19-2014--This approval includes revision of the three implementing Guides: DOE G 420.1-1, Nonreactor Nuclear Safety Design Criteria and Explosive Safety Criteria Guide for Use with DOE O 420.1, Facility Safety; DOE G 420.1-2, Guide for the Mitigation of Natural Phenomena Hazards for DOE Nuclear Facilities and NonNuclear Facilities; and DOE G 420.1-3, Implementation Guide for DOE Fire Protection and Emergency Services Programs for Use with DOE O 420.1B, Facility Safety

2010-09-23T23:59:59.000Z

427

What Has DOE Been Doing in the Area of Safety Culture?  

Broader source: Energy.gov [DOE]

Presenter: James Hutton, Acting Associate Deputy Assistant Secretary for Safety, Security, and Quality Programs and Chief Nuclear Safety Advisor Office of Environmental Management US Department of Energy

428

Tag: Safety  

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

8all en Best Practices Workshop for Safety Culture http:www.y12.doe.goveshbest-practices-workshop-safety-culture

429

Safety Information for Families  

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

Safety Information for Families Checking your home for hazards 22 safety items no home should be without Home Safety Checklists Helpful links Home Safety Council Hunter Safety:...

430

Generation IV Nuclear Energy Systems ...  

E-Print Network [OSTI]

Generation IV Nuclear Energy Systems ... The U.S. Department of Energy's Office of Nuclear Energy enhance safety and security, and develop nuclear power as an energy source for industrial applications Information ... U.S. Department of Energy www.energy.gov DOE Office of Nuclear Energy www.nuclear

Kemner, Ken

431

Biological Safety  

Broader source: Energy.gov [DOE]

The DOE's Biological Safety Program provides a forum for the exchange of best practices, lessons learned, and guidance in the area of biological safety. This content is supported by the Biosurety Executive Team. The Biosurety Executive Team is a DOE-chartered group. The DOE Office of Worker Safety and Health Policy provides administrative support for this group. The group identifies biological safety-related issues of concern to the DOE and pursues solutions to issues identified.

432

B PLANT DOCUMENTED SAFETY ANALYSIS  

SciTech Connect (OSTI)

This document provides the documented safety analysis (DSA) and Central Plateau Remediation Project (CP) requirements that apply to surveillance and maintenance (S&M) activities at the 221-B Canyon Building and ancillary support structures (B Plant). The document replaces BHI-010582, Documented Safety Analysis for the B-Plant Facility. The B Plant is non-operational, deactivated and undergoing long term S&M prior to decontamination and decommissioning (D&D). This DSA is compliant with 10 CFR 830, Nuclear Safety Management, Subpart B, ''Safety Basis Requirements.'' The DSA was developed in accordance with U.S. Department of Energy (DOE) standard DOE-STD-1120-98, Integration of Environment, Safety, and Health into Facility Disposition Activities (DOE 1998) per Table 2 of 10 CFR 830 Appendix A, DOE Richland Operation Office (RL) direction (02-ABD-0053, Fluor Hanford Nuclear Safety Basis Strategy and Criteria) for facilities in long term S&M, and RL Direction (02-ABD-0091, ''FHI Nuclear Safety Expectations for Nuclear Facilities in Surveillance and Maintenance''). A crosswalk was prepared to identify potential inconsistencies between the previous B Plant safety analysis and DOE-STD-1120-98 guidance. In general, the safety analysis met the criteria of DOE-STD-1120-98. Some format and content changes have been made, including incorporating recent facility modifications and updating the evaluation guidelines and control selection criteria in accordance with RL direction (02-ABD-0053). The facility fire hazard analysis (FHA) and Technical Safety Requirements (TSR) are appended to this DSA as an aid to the users, to minimize editorial redundancy, and to provide an efficient basis for update.

DODD, E.N.; KERR, N.R.

2003-08-01T23:59:59.000Z

433

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

and calculation of population exposures (resulting These were the largest reactorsreactor, cannot be since, the definitive, particularly ably contain many of the calculationsreactor safety study, WASH-1400 of General background and objective Summary of methodology and results WASH-1400 Accident sequence identification and calculation

Nero, A.V.

2010-01-01T23:59:59.000Z

434

Proceedings of the US Nuclear Regulatory Commission fourteenth water reactor safety information meeting: Volume 1, Plenary session, Severe accident sequence analysis, Risk analysis/PRA applications, Reference plant risk analysis - NUREG-1150, Innovative concepts for increased safety of advanced power reactors  

SciTech Connect (OSTI)

This six-volume report contains 156 papers out of the 175 that were presented at the Fourteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 27-31, 1986. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included thirty-four different papers presented by researchers from Canada, Czechoslovakia, Finland, Germany, Italy, Japan, Mexico, Spain, Sweden, Switzerland and the United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

Weiss, A.J. (comp.)

1987-02-01T23:59:59.000Z

435

NEW - DOE O 420.1 Chg 1, Facility Safety  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. This Page Change is limited in scope to changes necessary to invoke DOE-STD-1104, Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Document, and revised DOE-STD-3009-2014, Preparation of Nonreactor Nuclear Facility Documented Safety Analysis as required methods. DOE O 420.1C Chg 1, dated 2-27-15, cancels DOE O 420.1C, dated 12-4-12.

436

Safety | Argonne National Laboratory  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0 ResourceAwardsSafeguards and SecuritySafety for Users PrintSafety

437

Verifying Therapy Safety Interlock System with SPIN Motlatsi Seotsanyana1  

E-Print Network [OSTI]

, air traffic and space vehicle control systems, nuclear power plant safety control systems, medical as banking systems) to embedded environments (such as nuclear power plant safety control systemsVerifying Therapy Safety Interlock System with SPIN Motlatsi Seotsanyana1 , Jaco Geldenhuys2 1

Geldenhuys, Jaco

438

Louisiana Nuclear Energy and Radiation Control Law (Louisiana)  

Broader source: Energy.gov [DOE]

The Louisiana Department of Environmental Quality is responsible for the regulation of nuclear energy safety, permitting and radiation safety and control in Louisiana. The Department operates...

439

Recent developments in Topaz II reactor safety assessments  

SciTech Connect (OSTI)

In December 1991, the Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of a US launch of a Russian Topaz II space nuclear power system. The primary mission goal would be to demonstrate and evaluate Nuclear Electric Propulsion technology to establish a capability for future civilian and military missions. A preliminary nuclear safety assessment, involving selected safety analyses, was initiated to determine whether or not a space mission could be conducted safely and within budget constraints. This paper describes the preliminary safety assessment results and the nuclear safety program now being established for the Nuclear Electric Propulsion Space Test Program (NEPSTP).

Marshall, A.C.

1993-07-01T23:59:59.000Z

440

Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 2, Working Group Assessment Team reports; Vulnerability development forms; Working group documents  

SciTech Connect (OSTI)

The Secretary of Energy`s memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability.

Not Available

1993-11-01T23:59:59.000Z

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


441

ENVIRONMENTAL HEALTH & SAFETY EMPLOYEE SAFETY ORIENTATION  

E-Print Network [OSTI]

: FS Vancouver: Ops CHEMICAL SAFETY 27265 CONTRACTOR SAFETY 23867 EARLY RETURN TO WORK 23011 EMERGENCYENVIRONMENTAL HEALTH & SAFETY EMPLOYEE SAFETY ORIENTATION SIMON FRASER UNIVERSITY ENVIRONMENTAL HEALTH & SAFETY DEPARTMENT Discovery Park - MTF 8888 University Drive Burnaby, British Columbia Canada V5

442

NUCLEAR PLANT AND CONTROL  

E-Print Network [OSTI]

NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: software require- ments, safety analysis, formal for the digital protection systems of a nuclear power plant. When spec- ifying requirements for software and CRSA processes are described using shutdown system 2 of the Wolsong nuclear power plants as the digital

443

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.  

SciTech Connect (OSTI)

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are needed for repository modeling are severely lacking. In addition, most of existing reactive transport codes were developed for non-radioactive contaminants, and they need to be adapted to account for radionuclide decay and in-growth. The accessibility to the source codes is generally limited. Because the problems of interest for the Waste IPSC are likely to result in relatively large computational models, a compact memory-usage footprint and a fast/robust solution procedure will be needed. A robust massively parallel processing (MPP) capability will also be required to provide reasonable turnaround times on the analyses that will be performed with the code. A performance assessment (PA) calculation for a waste disposal system generally requires a large number (hundreds to thousands) of model simulations to quantify the effect of model parameter uncertainties on the predicted repository performance. A set of codes for a PA calculation must be sufficiently robust and fast in terms of code execution. A PA system as a whole must be able to provide multiple alternative models for a specific set of physical/chemical processes, so that the users can choose various levels of modeling complexity based on their modeling needs. This requires PA codes, preferably, to be highly modularized. Most of the existing codes have difficulties meeting these requirements. Based on the gap analysis results, we have made the following recommendations for the code selection and code development for the NEAMS waste IPSC: (1) build fully coupled high-fidelity THCMBR codes using the existing SIERRA codes (e.g., ARIA and ADAGIO) and platform, (2) use DAKOTA to build an enhanced performance assessment system (EPAS), and build a modular code architecture and key code modules for performance assessments. The key chemical calculation modules will be built by expanding the existing CANTERA capabilities as well as by extracting useful components from other existing codes.

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

2011-03-01T23:59:59.000Z

444

Technical evaluation report TMI action -- NUREG-0737 (II.D.1). Relief and safety valve testing, Watts Bar Nuclear Plant, Units 1 and 2 (Dockets 50-390 and 50-391)  

SciTech Connect (OSTI)

In the past, safety and relief valves installed in the primary coolant system of light water reactors have performed improperly. As a result, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) and, subsequently, NUREG-0737 (Clarification of TMI Action Plan Requirements) recommended development and completion of programs to do two things. First, they should reevaluate the functional performance capabilities of pressurized water reactor safety, relief, and block valves. Second, they should verify the integrity of the pressurizer safety and relief valve piping systems for normal, transient, and accident conditions. This report documents the review of those programs by Lockheed Idaho Technologies Company. Specifically, this report documents the review of the Watts Bar Nuclear Plant, Units 1 and 2, Applicant response to the requirements of NUREG-0578 and NUREG-0737. This review found the Applicant provided an acceptable response reconfirming they met General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 for the subject equipment. It should also be noted Lockheed Idaho performed this review for both Units 1 and 2. However, the applicability of this review to Unit 2 depends on verifying that the Unit 2 as-built system conforms to the Unit 1 design reviewed in this report.

Fineman, C.P.

1995-02-01T23:59:59.000Z

445

STUDENT HEALTH AND SAFETY TRAINING FORM Name: ______________________________ Department of ________________________  

E-Print Network [OSTI]

with Biohazardous agents Radiation Safety (nuclear) All who work with radioisotopes X-Ray Safety All who work with x-ray sources Laser Safety All who work with class IIIb and IV Lasers Specialized courses (by request) Ladder Safety For all who use ladders Fall Protection For all who work at heights with risk

Sinnamon, Gordon J.

446

Hot Cell Facility (HCF) Safety Analysis Report  

SciTech Connect (OSTI)

This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.

MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.; MAHN,JEFFREY A.; BERRY,DONALD T.; SCHWERS,NORMAN F.; VANDERBEEK,THOMAS E.; NAEGELI,ROBERT E.

2000-11-01T23:59:59.000Z

447

Bayesian network analysis of nuclear acquisitions  

E-Print Network [OSTI]

Nuclear weapons proliferation produces a vehement global safety and security concern. Perhaps most threatening is the scenario of a rogue nation or a terrorist organization acquiring nuclear weapons where the conventional ideas of nuclear deterrence...

Freeman, Corey Ross

2009-05-15T23:59:59.000Z

448

IMPORTANCE OF SAFETY CULTURE ASSESSMENT  

SciTech Connect (OSTI)

Safety Management has lately been considered by some Nuclear Regulatory agencies as the tool on which to concentrate their efforts to implement modern regulation structures, because Safety Culture was said to be difficult to monitor. However, Safety Culture can be assessed and monitored even if it is problematical to make Safety Culture the object of regulation. This paper stresses the feasibility and importance of Safety Culture Assessment based on self-assessment applications performed in several nuclear organizations in Latin America. Reasons and ownership for assessing Safety Culture are discussed, and relevant aspects considered for setting up and programming such an assessment are shown. Basic principles that were taken into account, as well as financial and human resources used in actual self-assessments are reviewed, including the importance of adequate statistical analyses and the necessity of proper feed-back of results. The setting up of action plans to enhance Safety Culture is the final step of the assessment program that once implemented will enable to establish a Safety Culture monitoring process within the organization.

Spitalnik, J.

2004-10-06T23:59:59.000Z

450

Adventures in scientific nuclear diplomacy  

SciTech Connect (OSTI)

A former director of Los Alamos National Laboratory offers a first-person perspective on the important contributions scientists can make toward improving the safety and security of nuclear materials and reducing the global nuclear dangers in an evolving world.

Hecker, Siegfried S. [Center for International Security and Cooperation, Stanford University, Stanford, California (United States)

2014-05-09T23:59:59.000Z

451

Implementation Guide for Use in Addressing Unreviewed Safety Question Requirements  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Guide provides information to assist in the implementation and interpretation of Title 10 Code of Federal Regulations (CFR) Part 830.203, Unreviewed Safety Question Process, of the Nuclear Safety Management Rules for applicable nuclear facilities owned or operated by the Department of Energy (DOE), including the National Nuclear Security Administration (NNSA). Canceled by DOE N 251.93.

2001-10-24T23:59:59.000Z

452

Implementation Guide for Use in Addressing Unreviewed Safety Question Requirements  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

This Guide provides information to assist in implementation of Title 10 Code of Federal Regulations Part 830.203, "Unreviewed Safety Question Process," of the Nuclear Safety Management Rules for Category 1, 2, and 3 nuclear facilities owned or operated by the Department of Energy, including the National Nuclear Security Administration. Cancels DOE G 424.1-1.

2006-07-24T23:59:59.000Z

453

Safety from the Operator's Perspective: We are All in this Together...  

Broader source: Energy.gov (indexed) [DOE]

Ellis President and CEO Institute of Nuclear Power Operations Jim Ellis President and CEO Institute of Nuclear Power Operations Safety from the Operator's Perspective: We are All...

454

Safety Values  

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

* Work-related injuries, illnesses and environmental incidents are preventable. * A just culture exists where safety and environmental concerns are brought forward without fear of...

455

Radiation Safety  

Broader source: Energy.gov (indexed) [DOE]

Weeks of training * 15 of that is OJT * General Code of Operating Rules * Air Brake & Train Handling * System Special Instructions * Safety Instructions * Federal Regulations *...

456

Implementation Guide for Use in Developing Technical Safety Requirements  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The guide was developed in support of Title 10 Code of Federal Regulations (CFR) Part 830, Nuclear Safety Management. It provides guidance for the requirements defined in 10 CFR §830.205, Technical Safety Requirements.

2014-09-08T23:59:59.000Z

457

Presentation, Safety from the Operator's Perspective: We are All in This Together, Safety Culture in All We Do  

Broader source: Energy.gov [DOE]

A presentation by Jim Ellis, President and CEO, Institute of Nuclear Power Operators (INPO), Safety from the Operator’s Perspective: We Are All in this Together, Safety Culture in All We Do.

458

System Design and the Safety Basis  

SciTech Connect (OSTI)

The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination & decommissioning (D&D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities.

Ellingson, Darrel

2008-05-06T23:59:59.000Z

459

INTEGRATED SAFETY MANAGEMENT SYSTEM SAFETY CULTURE IMPROVEMENT INITIATIVE  

SciTech Connect (OSTI)

In 2007, the Department of Energy (DOE) identified safety culture as one of their top Integrated Safety Management System (ISMS) related priorities. A team was formed to address this issue. The team identified a consensus set of safety culture principles, along with implementation practices that could be used by DOE, NNSA, and their contractors. Documented improvement tools were identified and communicated to contractors participating in a year long pilot project. After a year, lessons learned will be collected and a path forward determined. The goal of this effort was to achieve improved safety and mission performance through ISMS continuous improvement. The focus of ISMS improvement was safety culture improvement building on operating experience from similar industries such as the domestic and international commercial nuclear and chemical industry.

MCDONALD JA JR

2009-01-16T23:59:59.000Z

460

Nuclear Explosive and Weapon Surety Program - DOE Directives...  

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

1D Admin Chg 1, Nuclear Explosive and Weapon Surety Program by Carl Sykes Functional areas: Administrative Change, Defense Nuclear Facility Safety and Health Requirement, Defense...

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


461

Report to the DOE Nuclear Data Committee, 1982  

SciTech Connect (OSTI)

The report includes summaries of measurements and calculations of nuclear data applications, and use of nuclear data for reactor safety. (GHT)

Haght, R.C.; Struble, G.L.

1982-03-01T23:59:59.000Z

462

DOE Standard Integration Of Environment,Safety, and Health Into...  

Broader source: Energy.gov (indexed) [DOE]

of Energy (DOE) approved methodology for preparing a Documented Safety Analysis (DSA) for decommissioning of nuclear facilities, as well as environmental restoration activities...

463

Pantex Plant Achieves Key Safety Milestone Ahead of Schedule...  

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

Achieves Key Safety Milestone Ahead of Schedule | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing...

464

CRITICALITY SAFETY TRAINING AT FLUOR HANFORD (FH)  

SciTech Connect (OSTI)

The Fluor Hanford Criticality Safety engineers are extensively trained. The objectives and requirements for training are derived from Department of Energy (DOE) and American National Standards Institute/American Nuclear Society Standards (ANSI/ANS), and are captured in the Hanford Criticality Safety Program manual, HNF-7098. Qualification cards have been established for the general Criticality Safety Engineer (CSE) analyst, CSEs who support specific facilities, and for the facility Criticality Safety Representatives (CSRs). Refresher training and continuous education in the discipline are emphasized. Weekly Brown Bag Sessions keep the criticality safety engineers informed of the latest developments and historic perspectives.

TOFFER, H.

2005-05-02T23:59:59.000Z

465

Occupational Safety Review of High Technology Facilities  

SciTech Connect (OSTI)

This report contains reviews of operating experiences, selected accident events, and industrial safety performance indicators that document the performance of the major US DOE magnetic fusion experiments and particle accelerators. These data are useful to form a basis for the occupational safety level at matured research facilities with known sets of safety rules and regulations. Some of the issues discussed are radiation safety, electromagnetic energy exposure events, and some of the more widespread issues of working at height, equipment fires, confined space work, electrical work, and other industrial hazards. Nuclear power plant industrial safety data are also included for comparison.

Lee Cadwallader

2005-01-31T23:59:59.000Z

466

Setting clear expectations for safety basis development  

SciTech Connect (OSTI)

DOE-RL has set clear expectations for a cost-effective approach for achieving compliance with the Nuclear Safety Management requirements (10 CFR 830, Nuclear Safety Rule) which will ensure long-term benefit to Hanford. To facilitate implementation of these expectations, tools were developed to streamline and standardize safety analysis and safety document development resulting in a shorter and more predictable DOE approval cycle. A Hanford Safety Analysis and Risk Assessment Handbook (SARAH) was issued to standardized methodologies for development of safety analyses. A Microsoft Excel spreadsheet (RADIDOSE) was issued for the evaluation of radiological consequences for accident scenarios often postulated for Hanford. A standard Site Documented Safety Analysis (DSA) detailing the safety management programs was issued for use as a means of compliance with a majority of 3009 Standard chapters. An in-process review was developed between DOE and the Contractor to facilitate DOE approval and provide early course correction. As a result of setting expectations and providing safety analysis tools, the four Hanford Site waste management nuclear facilities were able to integrate into one Master Waste Management Documented Safety Analysis (WM-DSA).

MORENO, M.R.

2003-05-03T23:59:59.000Z

467

Department Safety Representatives Department Safety Representative  

E-Print Network [OSTI]

Department Safety Representatives Overview Department Safety Representative Program/Operations Guidance Document The Department Safety Representative (DSR) serves a very important role with implementation of safety, health, and environmental programs on campus. The role of the DSR is to assist

Pawlowski, Wojtek

468

Idaho National Laboratory Safety Presentations  

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

* Hand Tool Safety * Protect Your Hearing * Water Safety * Home Firearms Safety * Bicycle Safety * Pedestrian Safety * Others Outdoor Survival Safety (K-Middle School) What to...

469

Safety assured financial evaluation of maintenance  

E-Print Network [OSTI]

Management decisions in complex industrial facilities usually consider both the economic and environmental aspects of the plant's performance. For nuclear power plants (NPPs), safety is also a very substantial issue. The objectives...

Erguina, Vera

2004-09-30T23:59:59.000Z

470

Toolbox Safety Talk Hot Work Safety Procedures  

E-Print Network [OSTI]

Toolbox Safety Talk Hot Work Safety Procedures Environmental Health & Safety Facilities Safety-in sheet to Environmental Health & Safety for recordkeeping. "Hot Work" is defined as any temporary WORK Obtain a hot work permit from your supervisor or safety rep. Ensure fire/smoke detection

Pawlowski, Wojtek

471

Toolbox Safety Talk Machine Shop Safety  

E-Print Network [OSTI]

Toolbox Safety Talk Machine Shop Safety Environmental Health & Safety Facilities Safety & Health to Environmental Health & Safety for recordkeeping. Machine shops are an integral part of the Cornell University be taken seriously. Many of the most frequently cited OSHA safety standards pertain to machine safeguarding

Pawlowski, Wojtek

472

ENVIRONMENTAL HEALTH AND SAFETY GENERAL SAFETY MANUAL  

E-Print Network [OSTI]

ENVIRONMENTAL HEALTH AND SAFETY GENERAL SAFETY MANUAL May 10, 2002 #12;i Acknowledgements Environmental Health and Safety gratefully acknowledges the assistance provided by the University Safety Council extremely helpful. #12;ii Environmental Health and Safety General Safety Manual Table of Contents Section

Maroncelli, Mark

473

LASER SAFETY POLICY MANUAL ENVIRONMENTAL HEALTH & SAFETY  

E-Print Network [OSTI]

LASER SAFETY POLICY MANUAL ISSUED BY ENVIRONMENTAL HEALTH & SAFETY OFFICE OF RADIOLOGICAL SAFETY and GEORGIA TECH LASER SAFETY COMMITTEE July 1, 2010 Revised July 31, 2012 #12;Laser Safety Program 1-1 #12;Laser Safety Policy Manual TABLE OF CONTENTS 1. POLICY AND SCOPE

Houston, Paul L.

474

Safety, Security & Fire Report  

E-Print Network [OSTI]

2013 Safety, Security & Fire Report Stanford University #12;Table of Contents Public Safety About the Stanford University Department of Public Safety Community Outreach & Education Programs Emergency Access Transportation Safety Bicycle Safety The Jeanne Clery and Higher Education Act Timely Warning

Straight, Aaron

475

Perspective unconventional means for nuclear reactor control  

SciTech Connect (OSTI)

The concept of reliable shutdown of nuclear reactors demands application of engineering control means on the basis of principles, safety, safe failure, redundancy, independence, variety, defence in depth, and intrinsical safety. This report describes application of control methods.

Ionaitis, R.R.

1993-12-31T23:59:59.000Z

476

Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors  

E-Print Network [OSTI]

Safety. The Accident at TEPCO’s Fukushima Nuclear Power2: Accident and Thermal Fluids Analysis PIRTs. (Nuclearmolten nuclear reactor core debris following accidents such

Scarlat, Raluca Olga

2012-01-01T23:59:59.000Z

477

The Environmental Protection Agency's Safety Standards for Disposal of Spent Nuclear Fuel: Potential Path Forward in Response to the Report of the Blue Ribbon Commission on America's Nuclear Future - 13388  

SciTech Connect (OSTI)

Following the decision to withdraw the Yucca Mountain license application, the Department of Energy created a Blue Ribbon Commission (BRC) on America's Nuclear Future, tasked with recommending a national strategy to manage the back end of the nuclear fuel cycle. The BRC issued its final report in January 2012, with recommendations covering transportation, storage and disposal of spent nuclear fuel (SNF); potential reprocessing; and supporting institutional measures. The BRC recommendations on disposal of SNF and high-level waste (HLW) are relevant to the U.S. Environmental Protection Agency (EPA), which shares regulatory responsibility with the Nuclear Regulatory Commission (NRC): EPA issues 'generally applicable' performance standards for disposal repositories, which are then implemented in licensing. For disposal, the BRC endorses developing one or more geological repositories, with siting based on an approach that is adaptive, staged and consent-based. The BRC recommends that EPA and NRC work cooperatively to issue generic disposal standards-applying equally to all sites-early in any siting process. EPA previously issued generic disposal standards that apply to all sites other than Yucca Mountain. However, the BRC concluded that the existing regulations should be revisited and revised. The BRC proposes a number of general principles to guide the development of future regulations. EPA continues to review the BRC report and to assess the implications for Agency action, including potential regulatory issues and considerations if EPA develops new or revised generic disposal standards. This review also involves preparatory activities to define potential process and public engagement approaches. (authors)

Forinash, Betsy; Schultheisz, Daniel; Peake, Tom [U.S. Environmental Protection Agency, Radiation Protection Division (United States)] [U.S. Environmental Protection Agency, Radiation Protection Division (United States)

2013-07-01T23:59:59.000Z

478

The elements of nuclear power  

SciTech Connect (OSTI)

An introduction to the principles of nuclear fission power generation. Describes the physical processes which occur in a nuclear reactor and discusses the theory behind the calculations. Also covers heat transfer in reactors, thermodynamic power cycles, reactor operators, and radiation shielding. Material covered includes topics on the effects of nuclear radiation on humans, the safety of nuclear reactors and of those parts of the nuclear fuel cycle which deal with fuel element manufacture and the reprocessing of irradiated fuel.

Bennet, D.J.; Thomson, J.R.

1989-01-01T23:59:59.000Z

479

POWER PLANT RELIABILITY-AVAILABILITY AND STATE REGULATION. VOLUME 7 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Removal Equipment (nuclear plant) Turbine Building ClosedCooling Water System (nuclear plant) SteamReheater (nuclear plant) Inspection Water Induction

Nero, A.V.

2010-01-01T23:59:59.000Z

480

HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)  

SciTech Connect (OSTI)

The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S&M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard.

EVANS, C B

2004-12-21T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear safety bmu" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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481

Safety Hazards of Batteries  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0 ResourceAwardsSafeguards and Security Systems5, 2014Safety

482

Safety for Users  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0 ResourceAwardsSafeguards and Security Systems5,Updates-Safety for

483

Safety for Users  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0 ResourceAwardsSafeguards and Security Systems5,Updates-Safety

484

Safety for Users  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0 ResourceAwardsSafeguards and SecuritySafety for Users Print

485

Safety | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0 ResourceAwardsSafeguards and SecuritySafety for Users

486

Safety, Security & Environment  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0 ResourceAwardsSafeguards and SecuritySafety for

487

Perspectives on reactor safety  

SciTech Connect (OSTI)

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01T23:59:59.000Z

488

ENVIRONMENTAL HEALTH & SAFETY  

E-Print Network [OSTI]

ENVIRONMENTAL HEALTH & SAFETY ORIENTATION HANDBOOK Environmental Health and Safety Office safety & Safety Office 494-2495 (Phone) 494-2996 (Fax) Safety.Office@dal.ca (E-mail) www.dal.ca/safety (Web) Radiation Safety Office 494-1938 (Phone) 494-2996 (Fax) Melissa.Michaud@dal.ca (E-mail) University

Brownstone, Rob

489

Toolbox Safety Talk Welding & Metal Work Safety  

E-Print Network [OSTI]

Toolbox Safety Talk Welding & Metal Work Safety Environmental Health & Safety Facilities Safety or harmful emission giving metals. Welding Safety When welding outside of a designated welding booth, ensure injury. Avoid welding on materials such as galvanized or stainless steel in order to minimize toxic fume

Pawlowski, Wojtek

490

State-of-the-art of non-destructive testing methods and technologies for application to nuclear power plant safety-related concrete structures  

SciTech Connect (OSTI)

The inspection of nuclear power plant concrete structures presents challenges different from conventional civil engineering structures. Wall thicknesses can be in excess of one meter and the structures often have increased steel reinforcement density with more complex detailing. The accessibility for any testing method may be limited due to the presence of liners and other components and there can be a number of penetrations or cast-in-place items present. The objective of the report is to present the state-of-the art of non-destructive testing methods and technologies for the inspection of thick, heavily-reinforced nuclear power plant concrete cross-sections with particular respect to: •locating steel reinforcement and identification of its cover depth •locating tendon ducts and identification of the condition of the grout materials •detection of cracking, voids, delamination, and honeycombing in concrete structures •detection of inclusions of different materials or voids adjacent to the concrete side of the containment liner •methods capable of identification of corrosion occurrence on the concrete side of the containment liner.

Wiggenhauser, Herbert [BAM - Federal Institute for Materials Research and Testing, Berlin, D (Germany); Naus, Dan J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2014-02-18T23:59:59.000Z

491

State-of-the-Art of Non-Destructive Testing Methods and Technologies for Application to Nuclear Power Plant Safety-Related Concrete Structures  

SciTech Connect (OSTI)

The inspection of nuclear power plant concrete structures presents challenges different from conventional civil engineering structures. Wall thicknesses can be in excess of one meter and the structures often have increased steel reinforcement density with more complex detailing. The accessibility for any testing method may be limited due to the presence of liners and other components and there can be a number of penetrations or cast-in-place items present. The objective of the report is to present the state-of-the art of non-destructive testing methods and technologies for the inspection of thick, heavily-reinforced nuclear power plant concrete cross-sections with particular respect to: locating steel reinforcement and identification of its cover depth locating tendon ducts and identification of the condition of the grout materials detection of cracking, voids, delamination, and honeycombing in concrete structures detection of inclusions of different materials or voids adjacent to the concrete side of the containment liner methods capable of identification of corrosion occurrence on the concrete side of the containment liner

Wiggenhauser, Dr. Herbert [Federal Institute for Materials Research and Testing (BAM)] [Federal Institute for Materials Research and Testing (BAM); Naus, Dan J [ORNL] [ORNL

2014-01-01T23:59:59.000Z

492

Accident Response Group | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

involving nuclear weapons. The ARG staff includes scientists, engineers, technicians, health physics and safety specialist from NNSA's and the Department of Energy's national...

493

Gas Pipeline Safety (Indiana)  

Broader source: Energy.gov [DOE]

This section establishes the Pipeline Safety Division within the Utility Regulatory Commission to administer federal pipeline safety standards and establish minimum state safety standards for...

494

Electrical Safety Committee Charter  

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

ANL Electrical Safety Page DOE Electrical Safety Handbook General Statement Home & Office Equipment Statement APS Electrical Safety Update Guidelines for Working on Voltages < 240...

495

Safety Overview Committee (SOC)  

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

(SOC) Charter 1. Purpose The Safety Overview Committee establishes safety policies and ad hoc safety committees. 2. Membership Membership will include the following individuals:...

496

Nuclear Data Capabilities Supported by the DOE NCSP  

E-Print Network [OSTI]

Nuclear Data Capabilities Supported by the DOE NCSP Symposium on Nuclear Data for Criticality responsible for developing, implementing, and maintaining nuclear criticality safety. 3 #12;NCSP Technical the Production Codes and Methods for Criticality Safety Engineers (e.g. MCNP, SCALE, & COG) · Nuclear Data

Danon, Yaron

497

Asymptotic Safety  

E-Print Network [OSTI]

Asymptotic safety is a set of conditions, based on the existence of a nontrivial fixed point for the renormalization group flow, which would make a quantum field theory consistent up to arbitrarily high energies. After introducing the basic ideas of this approach, I review the present evidence in favor of an asymptotically safe quantum field theory of gravity.

R. Percacci

2008-11-18T23:59:59.000Z

498

Said Mughaghab Nuclear Data 2010,  

E-Print Network [OSTI]

IN THE UNRESOLVED ENERGY REGION FOR ENDF EVALUATIONS RPI Nuclear Data 2011 Symposium for Criticality SafetySaid Mughaghab Nuclear Data 2010, Jeju Island, Korea, April 26-30, 2010 ANALYSIS OF MEASUREMENTS and Reactor applications April 27, 2011 S. F. Mughabghab* National Nuclear Data Center Brookhaven National

Danon, Yaron

499

UNITED STATES NUCLEAR REGULATORY COMMISSION  

E-Print Network [OSTI]

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS ON RADIATION THERAPY OVEREXPOSURES IN PANAMA Addressees All medical licensees. Purpose The U.S. Nuclear times resulted in significant radiation overexposures to patients. The hospital staff did not perform

500

Temperature actuated automatic safety rod release  

DOE Patents [OSTI]

A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

Hutter, Ernest (Wilmette, IL); Pardini, John A. (Brookfield, IL); Walker, David E. (Naperville, IL)

1987-01-01T23:59:59.000Z