Licensed operating reactors

International Nuclear Information System (INIS)

Hartfield, R.A.

1990-03-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors

International Nuclear Information System (INIS)

1989-08-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1981-08-01

The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

Licensing of nuclear reactor operators

International Nuclear Information System (INIS)

1979-09-01

Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

Licensed operating reactors

International Nuclear Information System (INIS)

1990-01-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors

International Nuclear Information System (INIS)

1989-06-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units are provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Non-Power Reactor Operator Licensing Examiner Standards

International Nuclear Information System (INIS)

1994-06-01

The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR Part 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, this standard will be revised periodically to accommodate comments and reflect new information or experience

Licensing process for future applications of advanced-design nuclear reactors

International Nuclear Information System (INIS)

Miller, C.L.

1990-01-01

The existing 10CFR50 two-step licensing process in the Code of Federal Regulations can continue to be a viable licensing vehicle for future applications, at least for the near future. The US Nuclear Regulatory Commission (NRC) Commissioners and staff, the public, and the utilities (along with supporting architect/engineers and nuclear steam supply system vendors) have a vast body of experience and knowledge of the existing part 50 licensing process. All these participants are familiar with their respective roles in this process, and history shows this process to be a workable licensing vehicle. Nevertheless, the use of 10CFR52 should be encouraged for future applications. This proposed new rule is intended to achieve the early resolution of licensing issues, to reduce the complexity and uncertainty of the licensing process, and enhance the safety and reliability of nuclear power plants. Part 52's overall purpose is to improve reactor safety and streamline the licensing process by encouraging the use of standard reactor designs and by allowing the early resolution of site environmental and reactor safety issues. The public should be afforded an earlier entry into the licensing process as a result of design certification rulemaking process and combined construction permit/operating license hearings

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

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

Nuclear safety requirements for operation licensing of Egyptian research reactors

International Nuclear Information System (INIS)

Ahmed, E.E.M.; Rahman, F.A.

2000-01-01

From the view of responsibility for health and nuclear safety, this work creates a framework for the application of nuclear regulatory rules to ensure safe operation for the sake of obtaining or maintaining operation licensing for nuclear research reactors. It has been performed according to the recommendations of the IAEA for research reactor safety regulations which clearly states that the scope of the application should include all research reactors being designed, constructed, commissioned, operated, modified or decommissioned. From that concept, the present work establishes a model structure and a computer logic program for a regulatory licensing system (RLS code). It applies both the regulatory inspection and enforcement regulatory rules on the different licensing process stages. The present established RLS code is then applied to the Egyptian Research Reactors, namely; the first ET-RR-1, which was constructed and still operating since 1961, and the second MPR research reactor (ET-RR-2) which is now in the preliminary operation stage. The results showed that for the ET-RR-1 reactor, all operational activities, including maintenance, in-service inspection, renewal, modification and experiments should meet the appropriate regulatory compliance action program. Also, the results showed that for the new MPR research reactor (ET-RR-2), all commissioning and operational stages should also meet the regulatory inspection and enforcement action program of the operational licensing safety requirements. (author)

Nuclear reactor operator licensing

International Nuclear Information System (INIS)

Bursey, R.J.

1978-01-01

The Atomic Energy Act of 1954, which was amended in 1974 by the Energy Reorganization Act, established the requirement that individuals who had the responsibility of operating the reactors in nuclear power plants must be licensed. Section 107 of the act states ''the Commission shall (1) prescribe uniform conditions for licensing individuals; (2) determine the qualifications of such individuals; and (3) issue licenses to such individuals in such form as the Commission may prescribe.'' The article discusses the types of licenses, the selection and training of individuals, and the administration of the Nuclear Regulatory Commission licensing examinations

Lessons learned from the licensing process for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Dickson, P.W.; Clare, G.H.

1991-01-01

This paper presents the experience of licensing a specific liquid-metal fast breeder reactor (LMFBR), the Clinch River Breader Reactor Plant (CRBRP). It was a success story in that the licensing process was accomplished in a very short time span. The actions of the applicant and the actions of the US Nuclear Regulatory Commission (NRC) in response are presented and discussed to provide guidance to future efforts to license unconventional reactors. The history is told from the perspective of the authors. As such, some of the reasons given for success or lack of success are subjective interpretations. Nevertheless, the authors' positions provided them an excellent viewpoint to make these judgements. During the second phase of the licensing process, they were the CRBRP Technical Director and the Licensing Manager, respectively, for the Westinghouse Electric Corporation, the prime contractor for the reactor plant

Establishing a safety and licensing basis for generation IV advanced reactors. License by test

International Nuclear Information System (INIS)

Kadak, Andrew C.

2001-01-01

The license by test approach to licensing is a novel method of licensing reactors. It provides an opportunity to deal with innovative non-water reactors in a direct way on a time scale that could permit early certification based on tests of a demonstration reactor. The uncertainties in the design and significant contributors to risk would be identified in the PRA during the design. Deterministic analysis computer codes could be tested on a real reactor. Scaling effects and associated uncertainties would be minimized. License by test is an approach that has sufficient merit to be developed and tested

Operating reactors licensing actions summary. Volume 5, No. 2

International Nuclear Information System (INIS)

1985-04-01

The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the Operating Reactors Licensing Actions Program

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

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

Operating reactors licensing actions summary. Vol. 3, No. 3

International Nuclear Information System (INIS)

1983-04-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

International Harmonization of Reactor Licensing Regulations

International Nuclear Information System (INIS)

Kuhnt, Dietmar.

1977-01-01

The purpose of a harmonization policy for reactor licensing regulations on the basis of already considerable experience is to attain greater rationalisation in this field, in the interest of economic policy and healthy competition, and most important, radiation protection and safety of installations. This paper considers the legal instruments for such harmonization and the conditions for their implementation, in particular within the Communities framework. (NEA) [fr

Operating reactors licensing actions summary. Vol. 3, No. 6

International Nuclear Information System (INIS)

1983-07-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary. Volume 5, No. 6

International Nuclear Information System (INIS)

1985-08-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published for internal NRC use in managing the Operating Reactors Licensing Actions Program. Its content will change based on NRC management informational requirements

AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

International Nuclear Information System (INIS)

SUBUDHI, M.; MORANTE, R.; LEE, A.D.

2002-01-01

The reactor coolant system (RCS) mechanical components that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer. steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions. determination of the effects of aging on their intended safety functions. and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. In addition, this paper discusses time-limited aging analyses associated with neutron embrittlement of the reactor vessel beltline region and thermal fatigue

Operating reactors licensing actions summary. Volume 5, No. 7

International Nuclear Information System (INIS)

1985-09-01

The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Establishment of regulatory framework for the development reactor licensing

International Nuclear Information System (INIS)

Jo, Jong C.; Yune, Young G.; Kim, Woong S.; Ahn, Sang K.; Kim, In G.; Kim, Hho J.

2003-01-01

With a trend that various types of advanced reactor designs are currently under development worldwide, the Korea Atomic Energy Research Institute has been developing an advanced reactor called ' System-integrated Modular Advanced Reactor (SMART)', which is a small sized integral type pressurized water reactor with a rated thermal power of 330 MW. To demonstrate the safety and the performance of the SMART reactor design, the SMART Research and Development Center has embarked to build a scaled-down pilot plant of SMART, called 'SMART-P' with a rated thermal power of 65 MW. In preparation for the forthcoming applications for both construction permit and operating license of SMART-P in the near future, the Korea Institute of Nuclear Safety is developing a new regulatory framework for licensing review of such a development reactor, which covers establishment of licensing process, identification and resolution of technical and safety issues, development of regulatory evaluation or verification-purpose computer codes and analytical methods, and establishment of design-specific, general design and operating criteria, regulations, and associated regulatory guides. This paper presents the current activities for establishing a regulatory framework for the licensing of a research and development reactor. Discussions are made on the SMART-P development program, the current Korean regulatory framework for reactor licensing, the SMART-P licensing-related issues, and the approach and strategy for developing an effective regulatory framework for the SMART-P licensing

Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

International Nuclear Information System (INIS)

1995-06-01

The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience

Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

Operating reactors licensing actions summary. Volume 5, Number 1

International Nuclear Information System (INIS)

1985-03-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

Advanced CANDU reactor pre-licensing progress

International Nuclear Information System (INIS)

Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C.

2005-01-01

The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

Operating reactors licensing actions summary. Vol.4, No. 4

International Nuclear Information System (INIS)

1984-06-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

A BWR licensing experience in the USA

International Nuclear Information System (INIS)

Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

2015-09-01

The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

A BWR licensing experience in the USA

Energy Technology Data Exchange (ETDEWEB)

Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

2015-09-15

The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

Licensing of MAPLE reactors in Canada

International Nuclear Information System (INIS)

Lee, A.G.; Labrie, J.P.; Langman, V.J.

1999-01-01

Full text: The Operating Licence for a MAPLE reactor (i.e., a 10 MW(th), pool-type reactor), has been approved by the Atomic Energy Control Board (AECB) on August 16th, 1999. This Operating Licence has been obtained within three years of the initiation of the MDS Nordion Medical Isotopes Reactor (MMIR) project, which entails the design, construction and commissioning of two 10 MW MAPLE reactors at AECL's Chalk River Laboratories. The scope and nature of the information required by the AECB, the licensing process and highlights of the events which led to successfully obtaining the Operating Licence for the MAPLE reactor are discussed. These discussions address all phases of the licensing process (i.e., the environmental assessment in support of siting, the Preliminary Safety Analysis Report, PSAR, in support of design, procurement and construction, the Final Safety Analysis Report, FSAR, in support of commissioning and operations, and the development of suitable quality assurance subprograms for each phase). An overview of some of the unique technical aspects associated with the MAPLE reactors, and how they have been addressed during the licensing process are also provided (e.g., applying CSA N285.0, General Requirements for Pressure-Retaining Systems and Components in CANDU Nuclear Power Plants, to a small, low pressure, low temperature research reactor, confirmation of the performance of the driver fuel via laboratory and/or in-reactor testing, validation of the computer codes used to perform the safety analyses, critical parameter uncertainty assessment, full scale hydraulic testing of the performance of the design, fuel handling, human factors validation, operator training and certification). (author)

The strategy of experimental power reactor licensing in Indonesia

International Nuclear Information System (INIS)

Moch Djoko Birmano

2015-01-01

Currently, BATAN has being planned to develop Experimental Power Reactor (EPR), that is the research nuclear reactor that can generate power (electricity or heat). The EPR is planned will be built in the National Center for Research of Science and Technology (Puspiptek) area at Serpong, South Tangerang, Banten Province, with the choice of reactor types is HTGR with the power size of 10 MWth. As stated in the Act No. 10 year 1997 on Nuclear Power, that every construction and operation of nuclear reactors and other nuclear installations and decommissioning of nuclear reactors required to have a permit. Furthermore, the its implementation arrangements is regulated in Government Regulation (GR) No. 2 year 2014 on Licensing of Nuclear Installations and Nuclear Material Utilization, which contains the requirements and procedures for the licensing process since site, construction, commissioning, operation, and decommissioning, it means licensing is implemented during the activity of construction, operation and decommissioning of NPPs.While, for the more detailed licensing arrangements available in the guidelines of BAPETEN Chairman Regulation (BCR). This study was conducted to understand the legal and institutional aspects, types and stages, and the licensing process of RDE, and identify licensing strategy so that timely as planned. Methodologies used include the literature study, consultation with experts in BAPETEN, discussions in the national seminar including FGD. (author)

Systems engineering and the licensing of Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Kulesa, T., E-mail: tkulesa@us.ibm.com [IBM, Philidelphia, Pennsylvania (United States); Soderholm, K., E-mail: Kristiina.Soderholm@fortum.com [Fortum Power (Finland); Fechtelkotter, P., E-mail: pfech@us.ibm.com [IBM, Boston, Massacheusets (United States)

2014-07-01

Both global warming and the need for dependable sources of energy continue to make nuclear power generation an appealing option. But a history of cost overruns, project delays, and environmental disaster has pushed the industry to innovate and design a more flexible, scalable, and safe source of nuclear energy - the small modular reactor. Innovation in generation technology creates disruption in already complex licensing and regulatory processes. This paper discusses how the application of systems engineering and requirements management can help combat confusion, rework, and efficiency problems across the engineering and compliance life cycle. The paper is based on the PhD Dissertation 'Licensing Model Development for Small Modular Reactors (SMRs) - Focusing on Finnish Regulatory Framework', approved in 2013. The result of the study gives recommendations and tools to develop and optimize the licensing process for SMRs. The most important SMR-specific feature, in terms of licensing, is the modularity of the design. Here the modularity indicates multi-module SMR designs, which creates new challenges in the licensing process. Another feature impacting licensing feasibility is the plan to build many standardized power plants in series and use factory-fabricated modules to optimize the construction costs. SMR licensing challenges are under discussion in many international forums, such as World Nuclear Association Cooperation in Reactor Design Evaluation and Licensing Small Modular Reactor group (WNA CORDEL SMR) group and IAEA INPRO regulators' forum. This paper also presents an application of the new licensing process using Systems Engineering, Requirements Management, and Project Management practices and tools. (author)

Systems engineering and the licensing of Small Modular Reactors

International Nuclear Information System (INIS)

Kulesa, T.; Soderholm, K.; Fechtelkotter, P.

2014-01-01

Both global warming and the need for dependable sources of energy continue to make nuclear power generation an appealing option. But a history of cost overruns, project delays, and environmental disaster has pushed the industry to innovate and design a more flexible, scalable, and safe source of nuclear energy - the small modular reactor. Innovation in generation technology creates disruption in already complex licensing and regulatory processes. This paper discusses how the application of systems engineering and requirements management can help combat confusion, rework, and efficiency problems across the engineering and compliance life cycle. The paper is based on the PhD Dissertation 'Licensing Model Development for Small Modular Reactors (SMRs) - Focusing on Finnish Regulatory Framework', approved in 2013. The result of the study gives recommendations and tools to develop and optimize the licensing process for SMRs. The most important SMR-specific feature, in terms of licensing, is the modularity of the design. Here the modularity indicates multi-module SMR designs, which creates new challenges in the licensing process. Another feature impacting licensing feasibility is the plan to build many standardized power plants in series and use factory-fabricated modules to optimize the construction costs. SMR licensing challenges are under discussion in many international forums, such as World Nuclear Association Cooperation in Reactor Design Evaluation and Licensing Small Modular Reactor group (WNA CORDEL SMR) group and IAEA INPRO regulators' forum. This paper also presents an application of the new licensing process using Systems Engineering, Requirements Management, and Project Management practices and tools. (author)

Licensing of advanced reactors: Status report and perspective

International Nuclear Information System (INIS)

King, T.

1988-01-01

In July, 1986, the U.S. Nuclear Regulatory Commission issued a Policy State on the Regulation of Advanced Nuclear Power Plants. As part of this policy, advanced reactor designers were encouraged to interact with NRC [Nuclear Regulatory Commission] early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular high temperature gas-cooled reactor (MHTGR) and two liquid metal reactors (LMRs). This paper provides a status of the NRC review effort, describes the key policy and technical issues resulting from our review and provides the current status and approach to the development of licensing guidance on each

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1987-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1988-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicate but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: Nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat. (orig.)

Operating reactors licensing actions summary. Volume 5, No. 9

International Nuclear Information System (INIS)

1985-11-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Volume 5, No. 8

International Nuclear Information System (INIS)

1985-10-01

This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Volume 4, No. 9

International Nuclear Information System (INIS)

1984-11-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the division of licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Vol. 4, No. 2

International Nuclear Information System (INIS)

1984-04-01

This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

HTR-PM Safety requirement and Licensing experience

International Nuclear Information System (INIS)

Li Fu; Zhang Zuoyi; Dong Yujie; Wu Zongxin; Sun Yuliang

2014-01-01

HTR-PM is a 200MWe modular pebble bed high temperature reactor demonstration plant which is being built in Shidao Bay, Weihai, Shandong, China. The main design parameters of HTR-PM were fixed in 2006, the basic design was completed in 2008. The review of Preliminary Safety Analysis Report (PSAR) of HTR-PM was started in April 2008, completed in September 2009. In general, HTR- PM design complies with the current safety requirement for nuclear power plant in China, no special standards are developed for modular HTR. Anyway, Chinese Nuclear Safety Authority, together with the designers, developed some dedicated design criteria for key systems and components and published the guideline for the review of safety analysis report of HTR-PM, based on the experiences from licensing of HTR-10 and new development of nuclear safety. The probabilistic safety goal for HTR-PM was also defined by the safety authority. The review of HTR-PM PSAR lasted for one and a half years, with 3 dialogues meetings and 8 topics meetings, with more than 2000 worksheets and answer sheets. The heavily discussed topics during the PSAR review process included: the requirement for the sub-atmospheric ventilation system, the utilization of PSA in design process, the scope of beyond design basis accidents, the requirement for the qualification of TRISO coating particle fuel, and etc. Because of the characteristics of first of a kind for the demonstration plant, the safety authority emphasized the requirement for the experiment and validation, the PSAR was licensed with certain licensing conditions. The whole licensing process was under control, and was re-evaluated again after Fukushima accident to be shown that the design of HTR-PM complies with current safety requirement. This is a good example for how to license a new reactor. (author)

Challenges in licensing a sodium-cooled advanced recycling reactor

International Nuclear Information System (INIS)

Levin, Alan E.

2008-01-01

As part of the Global Nuclear Energy Partnership (GNEP), the U.S. Department of Energy (DOE) has focused on the use of sodium-cooled fast reactors (SFRs) for the destruction of minor actinides derived from used reactor fuel. This approach engenders an array of challenges with respect to the licensing of the reactor: the U.S. Nuclear Regulatory Commission (NRC) has never completed the review of an application for an operating license for a sodium-cooled reactor. Moreover, the current U.S. regulatory structure has been developed to deal almost exclusively with light-water reactor (LWR) designs. Consequently, the NRC must either (1) develop a new regulatory process for SFRs, or (2) reinterpret the existing regulations to apply them, as appropriate, to SFR designs. During the 1980s and 1990s, the NRC conducted preliminary safety assessments of the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Innovative Small Module (PRISM) designs, and in that context, began to consider how to apply LWR-based regulations to SFR designs. This paper builds on that work to consider the challenges, from the reactor designer's point of view, associated with licensing an SFR today, considering (1) the evolution of SFR designs, (2) the particular requirements of reactor designs to meet GNEP objectives, and (3) the evolution of NRC regulations since the conclusion of the SAFR and PRISM reviews. (author)

Licensed operating reactors

International Nuclear Information System (INIS)

1989-11-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the errata page

Licensed operating reactors status summary report data as of January 31, 1989

International Nuclear Information System (INIS)

Schwartz, I.

1989-03-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information, such as spent fuel storage capability, reactor years of experience, and non-power reactors in the United States

Licensed operating reactors: status summary report, data as of May 31, 1982

International Nuclear Information System (INIS)

1982-06-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of 02-28-89

International Nuclear Information System (INIS)

1989-03-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactor Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors status summary report, data as of December 31, 1988

International Nuclear Information System (INIS)

1989-02-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors. Status summary report data as of October 31, 1982

International Nuclear Information System (INIS)

1982-11-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors. Status summary report, data as of 8-31-82

International Nuclear Information System (INIS)

1982-09-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report data as of 9-30-86

International Nuclear Information System (INIS)

1987-03-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices. IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of 11-30-88

International Nuclear Information System (INIS)

1989-01-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information, such as spent fuel storage capability, reactor years of experience, and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of 11-30-86

International Nuclear Information System (INIS)

1987-06-01

The operating units status report - licensed operating reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, Data as of 08-31-86

International Nuclear Information System (INIS)

1987-03-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensing of the TRIGA Mark III reactor at the Mexican Nuclear Centre

International Nuclear Information System (INIS)

Ramirez, R.M.; Arrendondo, R.R.

1990-01-01

The TRIGA Mark III reactor at the Mexican Nuclear Centre went critical in 1968 and remained so until 1979 when the National Commission for Nuclear Safety and Safeguards (CNSNS), the Mexican regulatory authority, was set up. The reactor was therefore operating without a formal operating license, and the CNSNS accordingly requested the ININ to license the reactor under the existing conditions and to ensure that any modification of the original design complied with Standards ANSI/ANS-15 and with the code of practice set out in IAEA Safety Series No. 35. The most relevant points in granting the operating licence were: (a) the preparation of the Safety Report; (b) the formulation and application of the Quality Assurance Programme; (c) the reconditioning of the following reactor systems: the cooling systems; the ventilation and exhaust system; the monitoring system and control panel; (d) the training of the reactor operating staff at junior and senior levels; and (e) the formulation of procedures and instructions. Once the provisional operating license was obtained for the reactor it was considered necessary to modify the reactor core, which has been composed of 20% enriched standards fuel, to a mixed core based on a mixture of standard fuel and FLIP-type fuel with 70% 235 U enrichment. The CNSNS therefore requested that the mixed core be licensed and a technical report was accordingly annexed to the Safety Report, its contents including the following subjects: (a) neutron analysis of the proposed configuration; (b) reactor shutdown margins; (c) accident analysis; and (d) technical specifications. The licensing process was completed this year and we are now hoping to obtain the final operating license

Reactors licensing: proposal of an integrated quality and environment regulatory structure for nuclear research reactors in Brazil

International Nuclear Information System (INIS)

Serra, Reynaldo Cavalcanti

2014-01-01

A new integrated regulatory structure based on quality and integrated issues has been proposed to be implemented on the licensing process of nuclear research reactors in Brazil. The study starts with a literature review about the licensing process in several countries, all of them members of the International Atomic Energy Agency. After this phase it is performed a comparative study with the Brazilian licensing process to identify good practices (positive aspects), the gaps on it and to propose an approach of an integrated quality and environmental management system, in order to contribute with a new licensing process scheme in Brazil. The literature review considered the following research nuclear reactors: Jules-Horowitz and OSIRIS (France), Hanaro (Korea), Maples 1 and 2 (Canada), OPAL (Australia), Pallas (Holand), ETRR-2 (Egypt) and IEA-R1 (Brazil). The current nuclear research reactors licensing process in Brazil is conducted by two regulatory bodies: the Brazilian National Nuclear Energy Commission (CNEN) and the Brazilian Institute of Environment and Renewable Natural Resources (IBAMA). CNEN is responsible by nuclear issues, while IBAMA by environmental one. To support the study it was applied a questionnaire and interviews based on the current regulatory structure to four nuclear research reactors in Brazil. Nowadays, the nuclear research reactor’s licensing process, in Brazil, has six phases and the environmental licensing process has three phases. A correlation study among these phases leads to a proposal of a new quality and environmental integrated licensing structure with four harmonized phases, hence reducing potential delays in this process. (author)

A comparative analysis of the domestic and foreign licensing processes for power and non-power reactors

International Nuclear Information System (INIS)

Joe, J. C.; Youn, Y. K.; Kim, W. S.; Kim, H. J.

2003-01-01

The System-integrated Modular Advanced Reactor (SMART), a small to medium sized integral type Pressurized Water Reactor (PWR) has been developed in Korea. Now, SMART-P, a 1/5 scaled-down of the SMART, is being developed for the purpose of demonstrating the safety and performance of SMART design. The SMART-P is a first-of-a-kind reactor which is utilized for the research and development of a power reactor. Since the licensing process of such a reactor is not clearly specified in the current Atomic Energy Act, a comparative survey and analysis of domestic and foreign licensing processes for power and non-power reactors has been carried out to develop the rationale and technical basis for establishing the licensing process of such a reactor. The domestic and foreign licensing processes of power and non-power reactors have been surveyed and compared, including those of the U.S.A., Japan, France, U.K., Canada, and IAEA. The general trends in nuclear reactor classification, licensing procedures, regulatory technical requirements, and other licensing requirements and regulations have been investigated. The results of this study will be used as the rationale and technical basis for establishing the licensing process of reactors at development stage such as SMART-P

Risk-assessment techniques and the reactor licensing process

International Nuclear Information System (INIS)

Levine, S.

1979-01-01

A brief description of the Reactor Safety Study (WASH-1400), concentrating on the engineering aspects of the contribution to reactor accident risks is followed by some comments on how we have applied the insights and techniques developed in this study to prepare a program to improve the safety of nuclear power plants. Some new work we are just beginning on the application of risk-assessment techniques to stablize the reactor licensing process is also discussed

Aspects of the licensing procedures for enrichment reduction in research reactors

International Nuclear Information System (INIS)

Krull, W.

1983-01-01

The enrichment reduction for research reactors requires a licensing procedure. For this purpose the qualification of the new fuel has to be demonstrated and changes in reactor safety have to be investigated like reactivity values, form-factors, Pu- and fission product inventory, safety margins and accidents. Calculations should be partly experimentally verified. The possible extent of the licensing procedure is discussed. (orig.) [de

Licensing of ANSTO'S OPAL reactor during construction and commissioning

International Nuclear Information System (INIS)

Summerfield, M.W.; Ordonez, J.P.

2006-01-01

Full text: This paper presents a general description of the ongoing licensing activities associated with the construction and commissioning of the Australian Nuclear Science and Technology Organisation's (ANSTO) Open Pool Australian Light-water (OPAL) Reactor at their Lucas Heights site. It addresses the following aspects: The Construction Licence - what it is and what impact it had on the construction of the OPAL Reactor, specifically the various Construction Licence Conditions; The interface between ANSTO, INVAP and ARPANSA during the construction of the OPAL Reactor, particularly in relation to ARPANS Regulation 54; Specific licensing issues that have arisen during the construction and commissioning process and how they have been resolved; The Operating Licence Application - what it is and how it interfaces with the construction and commissioning of the OPAL Reactor. These aspects are all addressed from the point of view of the licensee ANSTO and the RRR Project. Particular emphasis will be given to the way in which the licensing process is integrated into the overall project program and the lessons learnt that may be of benefit to other licensees and regulators. Note that this paper is an update of a presentation given at IGORR10 and follows on from a paper previously presented at PNBC12 in October 2002

Set of rules SOR 2 licensing of nuclear reactors

International Nuclear Information System (INIS)

1976-05-01

This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

Science.gov (United States)

2013-01-28

... University of Illinois Advanced TRIGA Reactor, License No. R-115 The U.S. Nuclear Regulatory Commission (NRC) is noticing the termination of Facility Operating License No. R-115, for the University of Illinois... Operating License No. R-115 is terminated. The above referenced documents may be examined, and/or copied for...

The development of reactor operator license examination question bank

International Nuclear Information System (INIS)

Kim, In Hwan; Woo, S. M.; Kam, S. C.; Nam, K. J.; Lim, H. P.

2001-12-01

The number of NPP keeps increasing therefore there is more need of reactor operators. This trend requires the more efficiency in managing the license examination. Question bank system will help us to develop good quality examination materials and keep them in it. The ultimate purpose of the bank system is for selecting qualified reactor operators who are primarily responsible for the safety of reactor operation in NPP

Licensed operating reactors: Status summary report: Data as of February 29, 1988

International Nuclear Information System (INIS)

1988-04-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

International Nuclear Information System (INIS)

Lear, G.; Erickson, P.B.

1975-01-01

Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

Role of fission gas release in reactor licensing

International Nuclear Information System (INIS)

1975-11-01

The release of fission gases from oxide pellets to the fuel rod internal voidage (gap) is reviewed with regard to the required safety analysis in reactor licensing. Significant analyzed effects are described, prominent gas release models are reviewed, and various methods used in the licensing process are summarized. The report thus serves as a guide to a large body of literature including company reports and government documents. A discussion of the state of the art of gas release analysis is presented

Licensing process characteristics of Small Modular Reactors and spent nuclear fuel repository

Energy Technology Data Exchange (ETDEWEB)

Söderholm, Kristiina, E-mail: kristiina.soderholm@fortum.com [Fortum Power (Finland); Tuunanen, Jari, E-mail: jari.tuunanen@fortum.com [Fortum Power (Finland); Amaba, Ben, E-mail: baamaba@us.ibm.com [IBM Complex Systems (United States); Bergqvist, Sofia, E-mail: sofia.bergqvist@se.ibm.com [IBM Rational Software (Sweden); Lusardi, Paul, E-mail: plusardi@nuscalepower.com [NuScale Power (United States)

2014-09-15

Highlights: • We examine the licensing process challenges of modular nuclear facilities. • We compare the features of Small Modular Reactors and spent nuclear fuel repository. • We present the need of nuclear licensing simplification. • Part of the licensing is proposed to be internationally applicable. • Systems engineering and requirements engineering benefits are presented. - Abstract: This paper aims to increase the understanding of the licensing processes characteristics of Small Modular Reactors (SMR) compared with licensing of spent nuclear fuel repository. The basis of the SMR licensing process development lies in licensing processes used in Finland, France, the UK, Canada and the USA. These countries have been selected for this study because of their various licensing processes and recent actions in the new NPP construction. Certain aspects of the aviation industry licensing process have also been studied and selected practices have been investigated as possibly suitable for use in nuclear licensing. Suitable features for SMR licensing are emphasized and suggested. The licensing features of the spent nuclear fuel deep repository along with similar features of SMR licensing are discussed. Since there are similar types of challenges of lengthy licensing time frames, as well as modular features to be taken into account in licensing, these two different nuclear industry fields can be compared. The main SMR features to take into account in licensing are: • Standardization of the design. • Modularity. • Mass production. • Serial construction. Modularity can be divided into two different categories: the first category is simply a single power plant unit constructed of independently engineered modules (e.g. construction process for Westinghouse AP-1000 NPP) and the second one a power plant composed of many reactor modules, which are manufactured in factories and installed as needed (e.g. NuScale Power SMR design). The deep underground repository

Operating experience and maintenance at the TRIGA Mark II LENA reactor

International Nuclear Information System (INIS)

Cingoli, F.; Altieri, S.; Lana, F.; Rosti, G.; Alloni, L.; Meloni, S.

1988-01-01

The last two years at the Trigs Mark II LENA plant were characterized by the running of the n-n-bar oscillation NADIR experiment. Consequently reactor operation was positively affected and the running hours rose again above 1000 hours per year. The LENA team was also deeply involved in the procedures for the renewal of the reactor operation license. The new requirements set by the Nuclear Energy Licensing Authority (ENEA for Italy) most of which concerning radiation protection and environmental impact, have been already fulfilled. In some cases the installation of new apparatus is underway

Licensed operating reactors: Status summary report data as of June 30, 1988

International Nuclear Information System (INIS)

1988-07-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors. Status summary report, data as of April 30, 1982

International Nuclear Information System (INIS)

1982-05-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Management and Program Analysis, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

International Nuclear Information System (INIS)

Schwirian, R.; Robison, G.

1994-07-01

The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal

Licensed operating reactors

International Nuclear Information System (INIS)

Hartfield, R.A.

1994-03-01

The Nuclear Regulatory Commissions annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December, the year to date (in this case calendar year 1993) and cumulative data, usually for the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

Licensing schedule for away-from-reactor (AFR) spent fuel storage facilities

International Nuclear Information System (INIS)

Gray, P.L.

1981-08-01

The Nuclear Regulatory Commission has authority to issue licenses for Away-From-Reactor (AFR) installations for the storage of spent nuclear fuel. This report presents a detailed estimate of the time required to prosecute a licensing action. The projected licensing schedule shows that the elapsed time between filing an application and issuance of a license will be about 32 months, assuming intervention. The legal procedural steps will determine the time schedule and will override considerations of technical complexity. A license could be issued in about 14 months in the absence of intervention

Sodium Fast Reactor Safety and Licensing Research Plan

International Nuclear Information System (INIS)

Denman, Matthew; Lachance, Jeff; Sofu, Tanju; Wigeland, Roald; Flanagan, George; Bari, Robert

2013-01-01

Conclusions: The Sodium Fast Reactor Safety and Licensing Research Plan reports conclude a multi-year expert elicitation process. All information included in the studies are publicly available and the reports are UUR. These reports are intended to guide SFR researchers in the safety and licensing arena to important and outstanding issues Two (and a half) projects have been funded based on the recommendations in this report: • Modernization of SAS4A; • Incorporation of Contain/LMR with MELCOR; • (Data recovery at INL and PNNL)

Experience acquired by Furnas for licensing nuclear power plants

International Nuclear Information System (INIS)

Silva, A.J.C. da; Xavier, E.E.

1986-01-01

The system for licensing of Almirante Alvaro Alberto Nuclear Power Plant-Unit 1 is presented. The process phases for reactor construction and operation are described: preliminary site approval; bases for safety review; partial construction permits; final construction permits; emission of final report of safety analysis; initial operation license and permanent operation license. (M.C.K.) [pt

Cooperation in reactor design evaluation and licensing

International Nuclear Information System (INIS)

Kaufer, B.; Wasylyk, A.

2014-01-01

In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

Cooperation in reactor design evaluation and licensing

Energy Technology Data Exchange (ETDEWEB)

Kaufer, B.; Wasylyk, A. [World Nuclear Association, London (United Kingdom)

2014-07-01

In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

Licensed operating reactors: Status summary report, Data as of July 31, 1987

International Nuclear Information System (INIS)

1987-10-01

The Operating Units Status Report - Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, data as of 07-31-86

International Nuclear Information System (INIS)

1986-12-01

The Operating Units Status Report - Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management from the Headquarters staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, IE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: status summary report, data as of 6-30-81

International Nuclear Information System (INIS)

1981-07-01

The Operating Units Status Report - Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Management and Program Analysis from the Headquarters staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, IE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the U.S. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the U.S. energy situation as a whole

Licensed operating reactors: Status summary report, Data as of May 31, 1987

International Nuclear Information System (INIS)

1987-10-01

The OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, Data as of June 30, 1987

International Nuclear Information System (INIS)

1987-07-01

The Operating Units Status Report - Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, data as of 12-31-89

International Nuclear Information System (INIS)

1990-02-01

The Operating Units Status Report - Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, data as of September 30, 1989

International Nuclear Information System (INIS)

1989-12-01

THE OPERATING UNITS STATUS REPORT -- LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, data as of August 31, 1987

International Nuclear Information System (INIS)

1987-12-01

The Operating Units Status Report - Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, Data as of October 31, 1987

International Nuclear Information System (INIS)

Schwartz, I.

1987-12-01

The operating units status report - licensed operating reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, data as of 9-30-87

International Nuclear Information System (INIS)

1987-10-01

The OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, Data as of 2-28-87

International Nuclear Information System (INIS)

1987-10-01

The Operating Units Status Report- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors

International Nuclear Information System (INIS)

1991-08-01

The Nuclear Regulatory Commission's annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar 1990) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

Small modular reactors: current status, economic aspects and licensing

International Nuclear Information System (INIS)

Zimbron E, E.; Puente E, F.

2014-10-01

Interest for nuclear energy had resurgence since the beginning of the new century. This was a consequence of the new world conditions and needs: increasing energy demands (mainly from developing countries), awareness of the importance of energetic security and the necessity of limiting carbon emissions. In this nuclear boom, Small Modular Reactors (SMRs) develop and start to consolidate as a viable option for the present energy market. Their modular characteristics, lower initial capital cost, passive safety features and their niche applications, situate them as a technology with various advantages. The following study will present and analysis that will help to comprehend the SMRs present status. Information will show planned reactors, reactors in construction and in operation, advantages and challenges of their implementation, relevant economic aspects and important licensing factors that need to be highlighted. The analysis showed that the SMR technology is still in an initial stage that could reach and important development point in the next ten years. In this period, many of the reactors that are in design stage or that are through their licensing process might be constructed and could be getting ready for a commercial status. On the other hand, it has been observed that there are two main economic factors that need to be considered for any SMRs implementation project. First, the costs (initial, operation, maintenance, fuel and decommissioning) and second their possible niche market applications. Additionally, it has been noted that the licensing process is one of the greatest challenges for SMR general development. Licensing is mainly related to topic such as Emergency Planning Zone, first-of-a-kind engineering, passive safety features, proliferation resistance, multiple module designs and staffing. Previous information will serve as a base for carrying out a feasibility assessment analysis for SMR in Mexico. This part will be the last section of the project

Small modular reactors: current status, economic aspects and licensing

Energy Technology Data Exchange (ETDEWEB)

Zimbron E, E. [Instituto Tecnologico de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: erick.zimbron@gmail.com [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

Interest for nuclear energy had resurgence since the beginning of the new century. This was a consequence of the new world conditions and needs: increasing energy demands (mainly from developing countries), awareness of the importance of energetic security and the necessity of limiting carbon emissions. In this nuclear boom, Small Modular Reactors (SMRs) develop and start to consolidate as a viable option for the present energy market. Their modular characteristics, lower initial capital cost, passive safety features and their niche applications, situate them as a technology with various advantages. The following study will present and analysis that will help to comprehend the SMRs present status. Information will show planned reactors, reactors in construction and in operation, advantages and challenges of their implementation, relevant economic aspects and important licensing factors that need to be highlighted. The analysis showed that the SMR technology is still in an initial stage that could reach and important development point in the next ten years. In this period, many of the reactors that are in design stage or that are through their licensing process might be constructed and could be getting ready for a commercial status. On the other hand, it has been observed that there are two main economic factors that need to be considered for any SMRs implementation project. First, the costs (initial, operation, maintenance, fuel and decommissioning) and second their possible niche market applications. Additionally, it has been noted that the licensing process is one of the greatest challenges for SMR general development. Licensing is mainly related to topic such as Emergency Planning Zone, first-of-a-kind engineering, passive safety features, proliferation resistance, multiple module designs and staffing. Previous information will serve as a base for carrying out a feasibility assessment analysis for SMR in Mexico. This part will be the last section of the project

Licensed operating reactors status summary report data as of April 30, 1983. Vol. 7, No. 5

International Nuclear Information System (INIS)

1983-05-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: status summary report, data as of May 31, 1983. Volume 7, No. 6

International Nuclear Information System (INIS)

1983-06-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

International Nuclear Information System (INIS)

Beech, D.J.; May, R.

2000-01-01

In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

Licensing of the first reload of Angra-1 reactor

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1985-01-01

The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

Conversion to low-enriched fuel in research reactor aspects of licensing the Federal Republic of Germany

International Nuclear Information System (INIS)

Jacquemin, J.

1985-01-01

Conversion to low-enriched fuel and usage of new developed highly densified fuel in research-reactors will be an essential alteration in operating the reactor. According to the German Energy Act this has to be licensed. here might be some risk to the licensee of an older research-reactor by suspending his operating license because he cannot meet current requirements to be fulfilled or because of a court decision.Disposal of irradiated fuel elements of the new fuel type is a further significant problem which has to be solved before issuing a new license. (author)

Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Orvis, D.D.; Raabe, P.H.

1980-01-01

This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs

Licensing of the OPAL reactor during construction and commissioning

International Nuclear Information System (INIS)

Summerfield, M.

2007-01-01

This paper presents a description of the licensing activities associated with the construction and commissioning of the Australian Nuclear Science and Technology Organisation's (ANSTO) OPAL reactor. It addresses the Construction Licence, the interface between ANSTO, INVAP (the contractor with responsibility for design and construction of the facility) and the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA, the Australian nuclear regulator) during the construction of OPAL, specific licensing issues that have arisen during the construction and commissioning process, and the Operating Licence Application. Particular emphasis will be given to the way in which the licensing process is integrated into the overall project program and the lessons learnt that may be of benefit to other licensees and regulators

Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

International Nuclear Information System (INIS)

Carter, R.E.

1985-01-01

During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

Energy Technology Data Exchange (ETDEWEB)

Carter, R E

1985-07-01

During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

Safety-evaluation report related to renewal of the operating license for the Texas A and M University Research Reactor. Docket No. 50-128, License R-83

International Nuclear Information System (INIS)

1983-03-01

This Safety Evaluation Report for the application filed by the Texas A and M University (Texas A and M) for a renewal of operating license number R-83 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Texas Engineering and Experiment Station of the Texas A and M University and is located on the campus in College Station, Brazos County, Texas. The staff concludes that the TRIGA reactor facility can continue to be operated by Texas A and M University without endangering the health and safety of the public

Specificity in the licensing process of reduced enrichment in the Bulgarian research reactor

International Nuclear Information System (INIS)

Vitkova, Marietta; Gorinov, Ivan

2005-01-01

The presented paper considers some specific questions of the licensing process regarding the reconstruction of the Bulgarian IRT-2000 research reactor, which includes conversion to the low enriched fuel. This specificity has risen as a result of two facts. The design of the reactor reconstruction was made on the basis of the existing fresh 36% highly enriched fuel. But after finishing of the design process, this fresh highly enriched fuel was shipped back to Russia in the framework of the RERTR program. These facts have involved some changes in both - in the licensing and the design processes. Re-analysis of the neutronic and thermal-hydraulic calculations is required to be made on the base of the technical specifications of the new LEU fuel. To facilitate the licensing process the NRA has adopted regulatory acceptance criteria for approval of the reactor core design with LEU fuel. (author)

Licensed operating reactors: Status summary report, data as of 07-31-88

International Nuclear Information System (INIS)

1988-09-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of October 31, 1988

International Nuclear Information System (INIS)

1988-12-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the Errata page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of August 31, 1988

International Nuclear Information System (INIS)

1988-10-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of May 31, 1988

International Nuclear Information System (INIS)

Schwartz, I.

1988-07-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resource Management from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operation units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by the NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors status summary report, data as of April 30, 1989

International Nuclear Information System (INIS)

1989-06-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors. Status summary report, data as of 3-31-82

International Nuclear Information System (INIS)

1982-04-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Management and Program Analysis, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the Errata page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. II

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000 kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4 -8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear , plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Generic aging management programs for license renewal of BWR reactor coolant systems components

International Nuclear Information System (INIS)

Shah, V.N.; Liu, Y.Y.

2002-01-01

The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

International Nuclear Information System (INIS)

Shah, V.N.; Liu, Y.Y.

2002-01-01

The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

Improved best estimate plus uncertainty methodology, including advanced validation concepts, to license evolving nuclear reactors

International Nuclear Information System (INIS)

Unal, C.; Williams, B.; Hemez, F.; Atamturktur, S.H.; McClure, P.

2011-01-01

Research highlights: → The best estimate plus uncertainty methodology (BEPU) is one option in the licensing of nuclear reactors. → The challenges for extending the BEPU method for fuel qualification for an advanced reactor fuel are primarily driven by schedule, the need for data, and the sufficiency of the data. → In this paper we develop an extended BEPU methodology that can potentially be used to address these new challenges in the design and licensing of advanced nuclear reactors. → The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. → The methodology includes a formalism to quantify an adequate level of validation (predictive maturity) with respect to existing data, so that required new testing can be minimized, saving cost by demonstrating that further testing will not enhance the quality of the predictive tools. - Abstract: Many evolving nuclear energy technologies use advanced predictive multiscale, multiphysics modeling and simulation (M and S) capabilities to reduce the cost and schedule of design and licensing. Historically, the role of experiments has been as a primary tool for the design and understanding of nuclear system behavior, while M and S played the subordinate role of supporting experiments. In the new era of multiscale, multiphysics computational-based technology development, this role has been reversed. The experiments will still be needed, but they will be performed at different scales to calibrate and validate the models leading to predictive simulations for design and licensing. Minimizing the required number of validation experiments produces cost and time savings. The use of multiscale, multiphysics models introduces challenges in validating these predictive tools - traditional methodologies will have to be modified to address these challenges. This paper gives the basic aspects of a methodology that can potentially be used to address these new challenges in

The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

International Nuclear Information System (INIS)

Junkrans, S.; Vesterlund, G.; Vaernild, O.

1980-01-01

The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

Reactors Project Delivery: The Value of Experiance

International Nuclear Information System (INIS)

Stosic, V. Zoran

2014-01-01

State of Affairs: Energy Potential and Density versus Environmental Load of different Energy Sources, Development of Fuel into Energy/Electricity Generation, Production Costs of Electricity, Contributions of Nuclear Energy to Security of Energy Supply, Recent Nuclear Development, Public Support growing again. Projects Status: Reactors under Construction, Different Projects Industrial Schemes, Projects Overview. The Value of Experience: Licensing, Standardization on Early Engineering Activities, Supply Chain and Manufacturing of Heavy Components, Installation, Procurement. (author)

New trends in the evaluation and implementation of the safety-related operating experience associated with NRC-licensed reactors

International Nuclear Information System (INIS)

Michelson, C.; Heltemes, C.J.

1981-01-01

This article is an overview of the Nuclear Regulatory Commission program for the evaluation and dissemination of the safety-related operating experience associated with all NRC-licensed reactors. It discusses the historical background and past problems that led to the recent formation of NRC's Office for Analysis and Evaluation of Operational Data (AEOD) and details its activities, organization, staffing, and proposed analysis and evaluation methodology. The programs of industry organizations and nuclear plant licensees and the integration of foreign operating experience are included in the overview. The problems and limitations of the Licensee Event Report (LER) program and the Nuclear Plant Reliability Data system program are discussed. The AEOD analysis and evaluation methodology program includes some new improvements in the assessment of safety-related operating experience. Of particular note is the sequence coding and search procedure being developed by AEOD under a contract with the Nuclear Safety Information Center at the Oak Ridge National Laboratory. This computer-based retrieval system will have markedly improved search strategy capability for such items as commoncause failures or complex system interactions involving various failure sequences and other relationships associated with an event. The system retrieves failure data and information on the principal LER occurrence and on related component and system responses. The computer-generated Power Reactor Watch List enables AEOD to monitor all critical or unusual situations warranting close attention because of potential public health and safety. This listing is supported by a preestablished computer search strategy of the historical data base permitting identification of all past events and statistical information that are applicable to the situation being watched

Health physics aspects of advanced reactor licensing reviews

International Nuclear Information System (INIS)

Hinson, C.S.

1995-01-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on open-quotes next-generationclose quotes reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These open-quotes next-generationclose quotes reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four open-quotes next-generationclose quotes reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four open-quotes next-generationclose quotes reactor designs currently being reviewed by the NRC

Licensing procedures and safety criteria for research reactors in France

International Nuclear Information System (INIS)

Berry, J.L.; Lerouge, B.

1980-11-01

This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. Some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors. At the end, a few considerations are given to the consequences of the Osiris core conversion

77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

Science.gov (United States)

2012-05-03

... Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY: Nuclear Regulatory Commission... Commission (NRC or the Commission) has issued renewed Facility Operating License No. R- 112, held by Reed... License No. R-112 will expire 20 years from its date of issuance. The renewed facility operating license...

Licensing activities for the partial decommissioning of IRT-2000 research reactor in Sofia

International Nuclear Information System (INIS)

Apostolov, T.; Ilieva, Kr.; Papukchiev, A.; Kalchev, B.

2001-01-01

The project for refurbishment of IRT-2000 research reactor in Sofia into low-power reactor (200 kW) is based on the retention of some IRT-2000 buildings, facilities and equipment. The activities, which determine the partial decommissioning should be realized in accordance with preliminary developed licensing documents as General Plan, Safety Analysis Report and Environment Impact assessment Report. The goal of these documents is to provide and guarantee safe and effective activities with radioactive materials, to define strictly the dismantling procedures, and in the same time to minimize their influence on the environment. The Technical Tasks for General Plan, Safety Analysis Report and Environment Impact Assessment Report have been prepared and will be presented as preliminary licensing documents to the National Regulatory Body for approval before their application. A Quality Management system is being developed nowadays at INRNE. After its certification some requirements of the regulatory body will be completed. This certified QA system is a major part of the licensing procedure for the reconstruction of IRT-2000 research reactor. (author)

Licensed operating reactors. Status summary report as of February 29, 1984. Volume 8, No. 3

International Nuclear Information System (INIS)

1984-04-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement (IE), from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

The licensing processes influence on nuclear market

International Nuclear Information System (INIS)

Locatelli, Giorgio; Mancini, Mauro; Sainati, Tristano; Sallinen, Liisa

2011-01-01

The paper deals with the licensing nuclear power plants; it focuses primarily on the licensing process implications into the international nuclear market. Nowadays there are twenty-six countries that are planning to build new nuclear facilities, and thirty-seven where nuclear reactors are proposed; on the other hand, there are mainly ten international reactor vendors. At international level, there are few vendors that have sufficient resources, capabilities and experience to carry out the design and delivering of a nuclear power plant in the international market; On the other hand, the licensing processes are strictly dependent on national law frameworks, and on the nuclear policies. The paper proposes a comparison of six licensing processes (the ones established in Finland, France, Italy, South Korea, USA and UK), and analyzes its main features and implications; the IAEA licensing process is taken as reference point. The objective of the paper is to propose a systemic approach for considering the licensing procedures. The framework proposed enables facilitating the licensing management and inferring the main features of licensing contexts. The paper concludes with a forecast of the nuclear licensing context, especially with respect to the fourth generation of nuclear reactors. (author)

Health physics aspects of advanced reactor licensing reviews

Energy Technology Data Exchange (ETDEWEB)

Hinson, C.S. [Nuclear Regulatory Commission, Washington, DC (United States)

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.

NPP License Renewal and Aging Management: Revised Guidance

International Nuclear Information System (INIS)

Hull, A.B.; Hiser, A.L.; Lindo-Talin, S.E.

2012-01-01

Based on the Atomic Energy Act, the NRC issues licenses for commercial power reactors to operate for up to 40 years and allows these licenses to be renewed for up to another 20 years. NRC has approved license renewal (LR) for well over 50% of U.S. located reactors originally licensed to operate for 40 years. Of these 104 reactors (69 PWRs, 35 BWRs), the NRC has issued renewed licenses for 71 units and is currently reviewing applications for another 15 units. As of May 1, 2012, ten plants at nine sites had entered their 41st year of operation and thus are in their first period of extended operation (PEO). Five more plants will enter the PEO by the end of 2012. One foundation of the license renewal process has been license renewal guidance documents (LRGDs). The U.S. Nuclear Regulatory Commission (NRC) revised key guidance documents used for nuclear power LR in 2010 and 2011. These include NUREG-1800, 'Standard Review Plan for Review of License Renewal Applications,' revision 2 (SRP-LR), and NUREG-1801, 'Generic Aging Lessons Learned (GALL) Report,' revision 2 (GALL Report). The guidance documents were updated to reflect lessons learned and operating experience gained since the guidance documents were last issued in 2005. The reactor LRGDs referenced in this poster can all be accessed at http://www.nrc.gov/reactors/operating/licensing/renewal/guidance.html (author)

Licensing of the proposed PBMR-SA

International Nuclear Information System (INIS)

Clapisson, G.A.; Henderson, N.R.; Hill, T.F.; Keenan, N.H.; Metcalf, P.E.; Mysenkov, A.

1997-01-01

This paper describes the preliminary criteria, which are intended to be used by the South African regulatory authority (CNS), for licensing of the ESKOM proposed - South African high temperature gas-cooled Pebble Bed Modular Reactor (PBMR-SA). The CNS intends to apply the existing CNS licensing approach together with some international design criteria used for advanced reactors as well as international experience gained from the safety evaluation of the MHTGR, THTR, etc. A major requirement to this type of reactor is that it should comply with the current CNS risk criteria and provide, as a minimum, the same degree of protection to the operator, public and environment that is required for the current generation of nuclear reactors. (author)

Rules for the licensing of new experiments in BR2: application to the test irradiation of new MTR-fuels

International Nuclear Information System (INIS)

Joppen, F.

2000-01-01

New types of MTR fuel elements are being developed and require a qualification before routine operation could be authorized. During the test irradiation the new fuel elements .are considered as experimental devices and their irradiation is allowed according to the procedures for experiments. Authorization is based on the advice .of a consultative committee on experiments. This procedure is valid as long as the irradiation is covered by the actual reactor license. An additional license or an amendment is only required if due to the experiment the risk for the workers or the environment is increased in a significant way. A few experimental fuel plates loaded in the primary loop of the reactor will not increase this risk. The source term for potential radioactive releases remains more or less the same. The probability for an accident can be limited by restricting the heat flux and surface temperature. (author)

Licensing of ANSTO's Replacement Research Reactor

International Nuclear Information System (INIS)

Summerfield, M.W.; Garea, V.

2003-01-01

This paper presents a general description of the licensing of the 20 MW Pool-type Replacement Research Reactor (RRR) currently being built by the Australian Nuclear Science and Technology Organisation (ANSTO) at their Lucas Heights site. The following aspects will be addressed: 1) The influence of ARPANSA's (the Australian regulator) Regulatory Assessment Principles and Design Criteria on the design of the RRR. 2) The Site Licence Application, including the EIS and the supporting siting documentation. 3) The Construction Licence Application, including the PSAR and associated documentation. 4) The review process, including the IAEA Peer Review and the Public Submissions as well as ARPANSA's own review. 5) The interface between ANSTO, INVAP and ARPANSA in relation to the ongoing compliance with ARPANS Regulation 51 and 54. 6) The future Operating Licence Application, including the draft FSAR and associated documentation. These aspects are all addressed from the point of view of the licensee ANSTO and the RRR Project. Particular emphasis will be given to the way in which the licensing process is integrated into the overall project program and how the licensing and regulatory regime within Australia influenced the design of the RRR. In particular, the safety design features that have been incorporated as a result of the specific requirements of ANSTO and the Australian regulator will be briefly described. The paper will close with a description of how the RRR meets, and in many aspects exceeds the requirements of ANSTO and the Australian regulator. (author)

Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

International Nuclear Information System (INIS)

1978-09-01

The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas

Impact of the Three Mile Island accident on reactor safety and licensing in Canada

International Nuclear Information System (INIS)

Harvie, J.D.

1980-06-01

This paper discusses the implications of the accident at Three Mile Island on reactor safety and licensing in Canada. Reactor safety principles which can be learned from, or are reaffirmed by, the accident are reviewed. It is concluded that reactor safety demands a firm commitment to safety by all those involved in the nuclear industry. (auth)

License renewal of the HFR research reactor - A regulatory view

International Nuclear Information System (INIS)

Van der Plas, Y.

2006-01-01

Full text: Regulatory policy in The Netherlands is having safety evaluations of a nuclear installation each ten year. Recently a periodic safety evaluation has been finished on the 45 MW High Flux Reactor (HFR) in Petten divided over the following main topics: Risk Scoping Study; HEU-LEU fuel conversion; Evaluation against a pre-defined Reference Licensing Basis; Ageing study. The Risk Scoping Study is a limited probabilistic safety assessment showing scenarios towards serious core degradation and their probabilities. Individual risks for members of the public induced by the incidents analyzed in the first phase are compared with the mortality limits and probabilistic doses limits of the national BKSE decree. For non- proliferation reasons the fuel is converted from high to low enriched uranium. Before starting the project a reference licensing basis (RLB) was defined. A list of postulated initiating events (PIEs), re-establishing a design basis has been approved. An ageing management system shall provide information on the actual status and the residual life time of the plant. The evaluation results into a list of deficiencies, recommendations and suggestions from which modifications are to be proposed. Examples are: more redundancy and separation within the reactor protection system; renewal of the emergency and decay heat cooling; addition of a secundary reactor shut down system; more redundancy in the electric power supply; addition of injection valves introducing efficient use of passive cooling inventor) under LOCA conditions; damping of the impact from loads accidentally falling into the pool. A new nuclear license was issued begin 2004, covering all modifications which are to be implemented ultimate end 2007

Nuclear power reactor licensing and regulation in the United States

International Nuclear Information System (INIS)

Shapar, H.K.

1979-01-01

The report is devoted to four subjects: an explanation of the origins, statutory basis and development of the present regulatory system in the United States; a description of the various actions which must be taken by a license applicant and by the Nuclear Regulatory Commission before a nuclear power plant can be constructed and placed on-line, an account of the current regulatory practices followed by the US NRC in licensing nuclear power reactors; an identification of some of the 'lessons learned' from the Three Mile Island accident and some proposed regulatory and legislative solutions. (NEA) [fr

The SLOWPOKE licensing model

Energy Technology Data Exchange (ETDEWEB)

Snell, V. G.; Takats, F.; Szivos, K.

1989-08-15

The SLOWPOKE Energy System (SES-10) is a 10 MW heating reactor that has been developed in Canada. It will be capable of running without a licensed operator in continuous attendance, and will be sited in urban areas. It has forgiving safety characteristics, including transient time-scales of the order of hours. A process called `up-front` licensing has been evolved in Canada to identify, and resolve, regulatory concerns early in the process. Because of the potential market in Hungary for nuclear district heating, a licensing plan has been developed that incorporates Canadian licensing experience, identifies specific Hungarian requirements, and reduces the risk of licensing delays by seeking agreement of all parties at an early stage in the program.

Decentralization of operating reactor licensing reviews: NRR Pilot Program

International Nuclear Information System (INIS)

Hannon, J.N.

1984-07-01

This report, which has incorporated comments received from the Commission and ACRS, describes the program for decentralization of selected operating reactor licensing technical review activities. The 2-year pilot program will be reviewed to verify that safety is enhanced as anticipated by the incorporation of prescribed management techniques and application of resources. If the program fails to operate as designed, it will be terminated

77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

Science.gov (United States)

2012-11-15

... Research Institute TRIGA Reactor: Facility Operating License No. R-84 AGENCY: Nuclear Regulatory Commission... considering an application for the renewal of Facility Operating License No. R-84 (Application), which... the renewal of Facility Operating License No. R-84, which currently authorizes the licensee to operate...

Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

International Nuclear Information System (INIS)

Roudier, S.; Badel, D.; Beraha, R.; Champ, M.; Tricot, N.; Tran Dai, P.

1994-01-01

The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: 1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; 2) guidelines for nuclear design and manufacturing requirements related to safety and 3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs

Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

Energy Technology Data Exchange (ETDEWEB)

Roudier, S [Direction de la Surete des Installations Nucleaires, Fontenay-aux-Roses (France); Badel, D; Beraha, R [Direction Regionale de l` Industrie, de la Recherche et de l` Environnement Rhone-Alpes, Lyon (France); Champ, M; Tricot, N; Tran Dai, P [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

1994-12-31

The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: (1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; (2) guidelines for nuclear design and manufacturing requirements related to safety and (3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs.

NPP License Renewal and Aging Management: Revised Guidance

International Nuclear Information System (INIS)

Hull, A.B.; Hiser, A.L.; Lindo-Talin, S.E.

2012-01-01

Based on the Atomic Energy Act, the NRC issues licenses for commercial power reactors to operate for up to 40 years and allows these licenses to be renewed for up to another 20 years. NRC has approved license renewal for well over 50% of U.S. located reactors originally licensed to operate for 40 years. Of these 104 reactors (69 PWRs, 35 BWRs), the NRC has issued renewed licenses for 71 units and is currently reviewing applications for another 15 units. As of May 1, 2012, ten plants at nine sites had entered their 41st year of operation and thus are in their first period of extended operation (PEO). Five more plants will enter the PEO by the end of 2012. One foundation of the license renewal process has been license renewal guidance documents (LRGDs). The U.S. Nuclear Regulatory Commission (NRC) revised key guidance documents used for nuclear power license renewal in 2010 and 2011. These include NUREG-1800, 'Standard Review Plan for Review of License Renewal Applications,' revision 2 (SRP-LR), and NUREG-1801, 'Generic Aging Lessons Learned (GALL) Report,' revision 2 (GALL Report). The guidance documents were updated to reflect lessons learned and operating experience gained since the guidance documents were last issued in 2005. (author)

Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

International Nuclear Information System (INIS)

Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z.

2015-01-01

The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

Energy Technology Data Exchange (ETDEWEB)

Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z., E-mail: jonatasfmata@yahoo.com.br, E-mail: vasconv@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

Summary report on safety and licensing strategy support for the ABR prototype

International Nuclear Information System (INIS)

Cahalan, J. E.; Nuclear Engineering Division

2007-01-01

Argonne National Laboratory is providing support to the US Department of Energy in the Global Nuclear Energy Partnership (GNEP) in certification of an advanced, sodium-cooled fast reactor. The reactor is to be constructed as a prototype for future commercial power reactors that will produce electricity while consuming actinides recovered from light water reactor spent fuel. This prototype reactor has been called the Advanced Burner Reactor, or ABR, and is now often referred to as the advanced recycle reactor. As part of its activities, Argonne is providing technical services to assist definition of a safety and licensing strategy for the ABR prototype, and to further implementation of the strategy. In FY06, an organizational meeting was held for DOE and its laboratory contractors to discuss licensing alternatives and review previous licensing experience for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Plant (CRBRP). Near the end of FY06, a report summarizing the discussions and conclusions was written. One of the top-level conclusions recorded in the report was a recommendation to follow a licensing strategy that included the US Nuclear Regulatory Commission (NRC) as the regulatory review and licensing authority. In FY07, activities ar Argonne to support safety and licensing progress have continued. These activities have focused on further evaluation of licensing alternatives; assessment of design, analysis, and documentation implications of licensing paths; and initial technical interactions with the Nuclear Regulatory Commission. This report summarizes FY07 activities

Gas reactor international cooperative program interim report: United States/Federal Republic of Germany nuclear licensing comparison

International Nuclear Information System (INIS)

1978-09-01

In order to compare US and FRG Nuclear Licensing, a summary description of United States Nuclear Licensing is provided as a basis. This is followed by detailed information on the participants in the Nuclear Licensing process in the Federal Republic of Germany (FRG). FRG licensing procedures are described and the rules and regulations imposed are summarized. The status of gas reactor licensing in both the U.S. and the FRG is outlined and overall conclusions are drawn as to the major licensing differences. An appendix describes the most important technical differences between US and FRG criteria

The HOR core conversion program development and licensing experiences

International Nuclear Information System (INIS)

Vries, J.W. de; Gibcus, H.P.M.; Leege, P.F.A. de

1997-01-01

This paper deals with the experiences in the development of a fuel conversion program for a 2 MW university type research reactor, the HOR. It gives an overview of the technical and administrative aspects concerning the fuel conversion program development since the eighties, including the safety review and licensing process. The overall final safety report was submitted in 1995, together with the environmental impact report, and a licence application was submitted accordingly. The licence permitting the conversion was issued in 1996, coming into force at the beginning of this year, although an appeal case is still pending. At the moment the necessary preparations for starting the actual conversion of the HOR are made. The general program characteristics are addressed. (author)

Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

2007-03-01

The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

Assessment of specialized educational programs for licensed nuclear reactor operators

International Nuclear Information System (INIS)

Melber, B.D.; Saari, L.M.; White, A.S.; Geisendorfer, C.L.; Huenefeld, J.C.

1986-02-01

This report assesses the job-relatedness of specialized educational programs for licensed nuclear reactor operators. The approach used involved systematically comparing the curriculum of specialized educational programs for college credit, to academic knowledge identified as necessary for carrying out the jobs of licenses reactor operators. A sample of eight programs, including A.S. degree, B.S. degree, and coursework programs were studied. Subject matter experts in the field of nuclear operations curriculum and training determined the extent to which individual program curricula covered the identified job-related academic knowledge. The major conclusions of the report are: There is a great deal of variation among individual programs, ranging from coverage of 15% to 65% of the job-related academic knowledge. Four schools cover at least half, and four schools cover less than one-third of this knowledge content; There is no systematic difference in the job-relatedness of the different types of specialized educational programs, A.S. degree, B.S. degree, and coursework; and Traditional B.S. degree programs in nuclear engineering cover as much job-related knowledge (about one-half of this knowledge content) as most of the specialized educational programs

Legislative and regulatory aspects of nuclear power reactor licensing in the U.S.A

International Nuclear Information System (INIS)

Malsch, M.G.

1976-01-01

An explanation of the origins, statutory basis and development of the present regulatory system in the US. A description of the various actions which must be taken by a license applicant and by the USNRC before a nuclear power plant can be constructed and placed on-line. Account of the current regulatory practices followed by the USNRC in licensing nuclear power reactors. (orig./HP) [de

Sodium fast reactor safety and licensing research plan - Volume II

International Nuclear Information System (INIS)

Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

2012-01-01

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Licensing procedures and safety criteria for research reactors in France

International Nuclear Information System (INIS)

Berry, J.L.; Lerouge, B.

1983-01-01

From the very beginning of the CEA up to now, a great deal of work has been devoted to the development and utilization of research reactors in France for the needs of fundamental and applied research, production of radioisotopes, and training. In recent years, new reactors were commissioned while others were decommissioned. Moreover some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors (Osiris and Isis). This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. At the end, a few considerations are given to the consequences of the Osiris core conversion. Safety of research reactors has been studied in detail and many improvements have been brought due to: implementation of a specific experimental program, and adaptation of safety principles and rules elaborated for power reactors. Research reactors in operation in France have been built within a 22 year period. Meanwhile, safety rules have been improved. Old reactors do not comply with all the new rules but modifications are continuously made: after analysis of incidents, when replacement of equipment has to be carried out, when an important modification (fuel conversion for example) is decided upon

Sodium fast reactor safety and licensing research plan. Volume II.

Energy Technology Data Exchange (ETDEWEB)

Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

2012-05-01

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Licensing procedures and safety criteria for research reactors in France

Energy Technology Data Exchange (ETDEWEB)

Berry, J L; Lerouge, B [Centre d' Etudes Nucleaires de Saclay (France)

1983-08-01

From the very beginning of the CEA up to now, a great deal of work has been devoted to the development and utilization of research reactors in France for the needs of fundamental and applied research, production of radioisotopes, and training. In recent years, new reactors were commissioned while others were decommissioned. Moreover some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors (Osiris and Isis). This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. At the end, a few considerations are given to the consequences of the Osiris core conversion. Safety of research reactors has been studied in detail and many improvements have been brought due to: implementation of a specific experimental program, and adaptation of safety principles and rules elaborated for power reactors. Research reactors in operation in France have been built within a 22 year period. Meanwhile, safety rules have been improved. Old reactors do not comply with all the new rules but modifications are continuously made: after analysis of incidents, when replacement of equipment has to be carried out, when an important modification (fuel conversion for example) is decided upon.

NRC as referee (reactor licensing following the Three Mile Island accident)

International Nuclear Information System (INIS)

Eisenhut, D.G.

1984-01-01

In this article, the NRC's licensing director reports on the progress made by US utilities in complying with the key regulations stemming from the Three Mile Island accident. Over 130 items must be improved at more than 65 reactors. The actions taken by France in response to its own analysis of the accident are discussed. New NRC requirements with regard to operational safety, design, and emergency-response capability are outlined. Nearly all the training, or software, items in Nureg-0737 (''Clarification of TMI Action Plan Requirements'') and more than half of the mechanical, or hardware, items have been completed at plants with operating reactors. The Committee to Review Generic Requirements was created to develop means for controlling the number and nature of NRC requirements placed on licensees. Probabilistic risk-assessment techniques were not widely used by the NRC until after the Three Mile Island accident. The NRC has directed licensees and applicants for operating licenses to conduct control-room design reviews to identify and correct human-engineering discrepancies. Includes 2 tables

Experience in using a research reactor for the training of power reactor operators

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenaut, L.J.

1972-01-01

A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Format and Content. NUREG-1537, Part 1

International Nuclear Information System (INIS)

1996-02-01

NUREG - 1537, Part 1 gives guidance to non-power reactor licensees and applicants on the format and content of applications to the Nuclear Regulatory Commission for licensing actions. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination

Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Format and Content. NUREG-1537, Part 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-01

NUREG - 1537, Part 1 gives guidance to non-power reactor licensees and applicants on the format and content of applications to the Nuclear Regulatory Commission for licensing actions. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination.

Licensed operating reactors status summary report, data as of December 31, 1985. Volume 10, No. 1

International Nuclear Information System (INIS)

1986-01-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the Errata page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

Science.gov (United States)

2012-02-13

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-264; NRC-2012-0026] Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108 AGENCY: Nuclear Regulatory Commission... Facility Operating License No. R-108 (``Application''), which currently authorizes the Dow Chemical Company...

Future of nuclear licensing

International Nuclear Information System (INIS)

Denton, H.R.

1984-01-01

The following topics are outlined: Comparison of US and best foreign experience in nuclear power plant construction and operation; Status of licensing and construction; Observed attributes; Reduced construction time; Fewer reactor trips; Higher capacity factor; Diesel generator reliability; Steam generator tube leakage; and US regulatory initiatives: NRC efforts and industry efforts

Operator licensing examiner standards

International Nuclear Information System (INIS)

1993-01-01

The Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining licensees and applicants for reactor operator and senior reactor operator licenses at power reactor facilities pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). The Examiner Standards are intended to assist NRC examiners and facility licensees to better understand the initial and requalification examination processes and to ensure the equitable and consistent administration of examinations to all applicants. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator licensing policy changes

77 FR 42771 - License Renewal for the Dow Chemical TRIGA Research Reactor

Science.gov (United States)

2012-07-20

... Chemical Company in Midland, MI and is a part of the Analytical Sciences Laboratory. The reactor is housed... releases of liquid radioactive waste by any of the disposal methods. Solid low-level radioactive waste... addition, denial of the application for license renewal would cause the loss of the benefits of teaching...

Managing the high level waste nuclear regulatory commission licensing process

International Nuclear Information System (INIS)

Baskin, K.P.

1992-01-01

This paper reports that the process for obtaining Nuclear Regulatory Commission permits for the high level waste storage facility is basically the same process commercial nuclear power plants followed to obtain construction permits and operating licenses for their facilities. Therefore, the experience from licensing commercial reactors can be applied to the high level waste facility. Proper management of the licensing process will be the key to the successful project. The management of the licensing process was categorized into four areas as follows: responsibility, organization, communication and documentation. Drawing on experience from nuclear power plant licensing and basic management principles, the management requirement for successfully accomplishing the project goals are discussed

Licensing procedures and safety criteria for core conversion in Japan

International Nuclear Information System (INIS)

Kanda, K.; Nakagome, Y.; Hayashi, M.

1983-01-01

In Japan, the establishment and operation of nuclear installations are governed mainly by the Law for Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors. This law lays down the regulations and conditions for licensing of the various installations involved in the nuclear fuel cycle, namely licensing of installations for refining, fabricating and reprocessing; and reactors, as well as licensing of the use of nuclear fuels in research facilities. Although procedures for the installations listed above vary depending on the installation concerned, only those relating to construction and operation of reactor facilities will be analysed in this study, as the conditions and principles applying to licensing and control of other installations are, to a large extent, similar to those concerning reactor facilities. The second part of this presentation describes the safety review of the KUCA reactor core conversion form HEU to MEU. For the safety review of the core conversion, the Committee on Examination of Reactor Safety of Japanese Government examined mainly the the nuclear characteristics and the integrity of aluminide fuel plates, which was very severe because we had no experience to use aluminide fuel plates in Japan. The integrity of fuel plates and the results of the worst accident analysis for the MEU core are shown with the comparison between the HEU and MEU cores. The significant difference was not observed between them. All the regulatory procedures were completed in September 1980. Fabrication of MEU fuel elements for the KUCA experiments by CERCA in France was started in September 1980, and will be completed in March 1981. The critical experiments in the KUCA with MEU fuel will be started on a single-core in May 1981 as a first step. Those on a coupled-core will follow

Undergraduate reactor control experiment

International Nuclear Information System (INIS)

Edwards, R.M.; Power, M.A.; Bryan, M.

1992-01-01

A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

License renewal process

International Nuclear Information System (INIS)

Fable, D.; Prah, M.; Vrankic, K.; Lebegner, J.

2004-01-01

The purpose of this paper is to provide information about license renewal process, as defined by Nuclear Regulatory Commission (NRC). The Atomic Energy Act and NRC regulations limit commercial power reactor licenses to an initial 40 years but also permit such licenses to be renewed. This original 40-year term for reactor licenses was based on economic and antitrust considerations not on limitations of nuclear technology. Due to this selected time period; however, some structures and components may have been engineered on the basis of an expected 40-year service life. The NRC has established a timely license renewal process and clear requirements codified in 10 CFR Part 51 and 10 CFR Part 54, that are needed to assure safe plant operation for extended plant life. The timely renewal of licenses for an additional 20 years, where appropriate to renew them, may be important to ensuring an adequate energy supply during the first half of the 21st Century. License renewal rests on the determination that currently operating plants continue to maintain adequate levels of safety, and over the plant's life, this level has been enhanced through maintenance of the licensing bases, with appropriate adjustments to address new information from industry operating experience. Additionally, NRC activities have provided ongoing assurance that the licensing bases will continue to provide an acceptable level of safety. This paper provides additional discussion of license renewal costs, as one of key elements in evaluation of license renewal justifiability. Including structure of costs, approximately value and two different approaches, conservative and typical. Current status and position of Nuclear Power Plant Krsko, related to license renewal process, will be briefly presented in this paper. NPP Krsko is designed based on NRC Regulations, so requirements from 10 CFR 51, and 10 CFR 54, are applicable to NPP Krsko, as well. Finally, this paper will give an overview of current status of

Experiences with interpretation and application of the German Atomic Energy Act for the German research reactors in Geesthacht FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1980-01-01

The German research reactors FRG-1 and FRG-2 have passed different types of licensing procedures in the past years. It is reported about the experiences we have got in the interpretation and application of section 7 of the German Atomic Energy Act. Following these experiences an estimation is done for the licensing procedure for the reduction of the uranium enrichment. (orig.) [de

Licensing of HTGRs in the United States

International Nuclear Information System (INIS)

Fisher, C.R.; Orvis, D.D.

1981-01-01

The licensing history of the high-temperature gas-cooled reactor (HTGR) in the United States is given historical perspective. The experience began with the licensing of the Peach Bottom Atomic Power Station and extends to the continuing experience at the Fort St. Vrain Nuclear Generating Station. Additional experience was obtained from the licensing reviews in the mid-1970s of the large HTGR plants that were to be built by Philadelphia Electric Company and Delmarva Power and Light. Also, information was provided by the licensing review of the General Atomic standard plant by the U.S. Nuclear Regulatory Commission (NRC) at about the same time. These experiences are summarized in terms of the principal design criteria that were required by the regulatory authority for each project. These criteria include specification of the design basis accidents that were postulated for the plant safety analysis. Several technical issues raised by the NRC during their review of the large HTGR are presented. (author)

Licensing of HTGRs in the United States

Energy Technology Data Exchange (ETDEWEB)

Fisher, C. R.; Orvis, D. D. [General Atomic Co., San Diego, CA (USA)

1981-01-15

The licensing history of the high-temperature gas-cooled reactor (HTGR) in the United States is given historical perspective. The experience began with the licensing of the Peach Bottom Atomic Power Station and extends to the continuing experience at the Fort St. Vrain Nuclear Generating Station. Additional experience was obtained from the licensing reviews in the mid-1970s of the large HTGR plants that were to be built by Philadelphia Electric Company and Delmarva Power and Light. Also, information was provided by the licensing review of the General Atomic standard plant by the U.S. Nuclear Regulatory Commission (NRC) at about the same time. These experiences are summarized in terms of the principal design criteria that were required by the regulatory authority for each project. These criteria include specification of the design basis accidents that were postulated for the plant safety analysis. Several technical issues raised by the NRC during their review of the large HTGR are presented.

10 CFR 110.26 - General license for the export of nuclear reactor components.

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF... Denmark Finland France Germany Greece Indonesia Ireland Italy Japan Latvia Lithuania Luxembourg...

Reactor safety

International Nuclear Information System (INIS)

Butz, H.P.; Heuser, F.W.; May, H.

1985-01-01

The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

International Nuclear Information System (INIS)

1996-02-01

NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination

Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-01

NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination.

Reactor coolant pump service life evaluation for current life cycle optimization and license renewal

International Nuclear Information System (INIS)

Doroshuk, B.W.; Berto, D.S.; Robles, M.

1990-01-01

This paper reports that as part of the plant life cycle management and license renewal program, Baltimore Gas and Electric Company (BG and E) has completed a service life evaluation of their reactor coolant pumps, funded jointly by EPRI and performed by ABB Combustion Engineering Nuclear Power. Two of the goals of the BG and E plant life cycle management and license renewal program, and of this current evaluation, are to identify actions which would optimize current plant operation, and ensure that license renewal remains a viable option. The reactor coolant pumps (RCPs) at BG and E's Calvert Cliffs Units 1 and 2 are Byron Jackson pumps with a diffuser and a single suction. This pump design is also used in many other nuclear plants. The RCP service life evaluation assessed the effect of all plausible age-related degradation mechanisms (ARDMs) on the RCP components. Cyclic fatigue and thermal embrittlement were two ARDMs identified as having a high potential to limit the service life of the pump case. The pump case is a primary pressure boundary component. Hence, ensuring its continued structural integrity is important

The regulation and licensing of research reactors and associated facilities in the United Kingdom

International Nuclear Information System (INIS)

Weightman, M.W.; Willby, C.R.

1990-01-01

In the United Kingdom, the Nuclear Installations Inspectorate (NII) licenses nuclear facilities, including research reactors, on behalf of the Health and Safety Executive (HSE). The legislation, the regulatory organizations and the methods of operation that have been developed over the last 30 years result in a largely non-prescriptive form of control that is well suited to research reactors. The most important part of the regulatory system is the license and the attachment of conditions which it permits. These conditions require the licensee to prepare arrangements to control the safety of the facility. In doing so the licensee is encouraged to develop a 'safety culture' within its organization. This is particularly important for research reactors which may have limited staff resources and where the ability, and at times the need, to have access to the core is much greater than for nuclear power plants. Present day issues such as the ageing of nuclear facilities, public access to the rationale behind regulatory decisions, and the emergence of more stringent safety requirements, which include a need for quantified safety criteria, have been addressed by the NII. This paper explores the relevance of such issues to the regulation of research reactors. In particular, it discusses some of the factors associated with research reactors that should be considered in developing criteria for the tolerability of risk from these nuclear facilities. From a consideration of these factors, it is the authors' view that the range of tolerable risk to the public from the operation of new research reactors may be expected to be more stringent than similar criteria for new nuclear power plants, whereas the criteria for tolerable risk for research reactor workers are expected to be about the same as those for power reactor workers

Two Approaches to Reactor Decommissioning: 10 CFR Part 50 License Termination and License Amendment, Lessons Learned from the Regulatory Perspective

International Nuclear Information System (INIS)

Watson, B.A.; Buckley, J.T.; Craig, C.M.

2006-01-01

Trojan Nuclear Plant (Trojan) and Maine Yankee Nuclear Plant (Maine Yankee) were the first two power reactors to complete decommissioning under the U. S. Nuclear Regulatory Commission's (NRC's) License Termination Rule (LTR), 10 CFR Part 20, Subpart E. The respective owners' decisions to decommission the sites resulted in different approaches to both the physical aspects of the decommissioning, and the approach for obtaining approval for completing the decommissioning in accordance with regulations. Being in different States, the two single-unit pressurized water reactor sites had different State requirements and levels of public interest that impacted the decommissioning approaches. This resulted in significant differences in decommissioning planning, conduct of decommissioning operations, volumes of low- level radioactive waste disposed, and the final status survey (FSS) program. While both licensees have Independent Spent Fuel Storage Installations (ISFSIs), Trojan obtained a separate license for the ISFSI in accordance with the requirements of 10 CFR Part 72 and terminated their 10 CFR Part 50 license. Maine Yankee elected to obtain a general license under 10 CFR Part 50 for the ISFSI and reduce the physical site footprint to the ISFSI through a series of license amendments. While the NRC regulations are flexible and allow different approaches to ISFSI licensing there are separate licensing requirements that must be addressed. In 10 CFR 50.82, the NRC mandates public participation in the decommissioning process. For Maine Yankee, public input resulted in the licensee entering into an agreement with a concerned citizen group and resulted in State legislation that significantly lowered the dose limit below the NRC radiological criteria of 25 mrem (0.25 mSv) per year (yr) in 10 CFR 20.1402 for unrestricted use. The lowering of the radiological criteria resulted in a significant dose modeling effort using site-specific Derived Concentrations Guideline Levels (DCGLs

Licensing Support Experience of the BN-600 Operation

International Nuclear Information System (INIS)

Khrennikov, N.; Sintsov, A.

2013-01-01

License procedure - Main principle: • All works, including fatigue tests of new types of fuel, are carried out at the unit 3 Beloyarsk nuclear power plants with the BN-600 reactor with the justification of the regulatory body. • Justification procedure is standard for all power units and independent from the reactor types. • The regulatory body and independent experts or technical support organizations, which can be involved in this work by the regulatory body, review SAR, operational manuals and other operator documents. • Safety requirements (i.e. Federal rules and codes). The project and design documents shall meet safety requirements. • The technical and organizational measures for safety guarantee shall meet well-known results of the research investigations or shall be experimental validate

Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

International Nuclear Information System (INIS)

1977-06-01

ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

Simulator experiments: effects of experience of senior reactor operators and of presence of a shift technical advisor on performance in a boiling water reactor control room

International Nuclear Information System (INIS)

Beare, A.N.; Dorris, R.E.; Gray, L.H.

1984-12-01

This report describes the first experiment in a Nuclear Regulatory Commission-sponsored program of training simulator experiments and field data collection to evaluate the effects of selected performance shaping factors on the performance of nuclear power plant control room operators. The factors investigated were the experience level of the Senior Reactor Operator (SRO) and the presence of a Shift Technical Advisor (STA). Data were collected from 16 two-man crews of licensed operators (one SRO and one RO). The crews were split into high and low SRO-experience groups on the basis of the years of experience of the SROs as SROs. One half (4 of the 8 crews in each group) of the high- and low-SRO experience groups were assisted by an STA or an SRO acting as an STA. The crews responded to four simulated plant casualties which ranged in severity from an uncomplicated turbine trip to an anticipated transient without scram (ATWS). No significant differences in overall performance were found between groups led by high (25 to 114 months licensed as an SRO) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have shorter task performance times than crews led by high experience SROs. Although a tendency for the STA-assisted groups to score higher on four of the five measures was observed, the presence of the STA had no statistically significant effect on overall team performance. The correlation between individual performance, as measured by four of the task performance measures, and experience, measured by months as a licensed operator, was not statistically significant, nor was the correlation between task performance and recency of simulator training. 18 references, 5 figures, 13 tables

Scaling analysis in bepu licensing of LWR

International Nuclear Information System (INIS)

D'auria, Francesco; Lanfredini, Marco; Muellner, Nikolaus

2012-01-01

'Scaling' plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a 'roadmap' to scaling. Key elements are the 'scaling-pyramid', related 'scaling bridges' and a logical path across scaling achievements (which constitute the 'scaling puzzle'). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the 'key-to-scaling', in the licensing process of a nuclear power plant. The proposed 'road map to scaling' aims at solving the 'scaling puzzle', by introducing a unified approach to the problem.

IRIS pre-application licensing

International Nuclear Information System (INIS)

Carelli, Mario D.; Kling, Charles L.; Ritterbusch, Stanley E.

2003-01-01

This paper presents the approach to pre-application licensing by the International Reactor Innovative and Secure (IRIS), and advanced, integral reactor design with a thermal power of 1000 MW. The rationale for the pre-application licensing is discussed. Since IRIS technology is based on proven LWR experience, the project will rely on AP600/AP1000 precedent and will focus during the pre-application on long lead and novel items. A discussion of the evolution of the project to significantly reduce licensing issues is provided, followed by a summary of the IRIS safety-by-design which provides a formidable first step in the Defense in Depth approach. The effects of the safety-by-design, as well as of passive systems, on the IRIS safety will be investigated in a proposed testing program that will be reviewed by NRC during the pre-application. Documentation to be provided to NRC is discussed. Early design analyses indicate that the benefits of the IRIS safety-by-design approach are so significant that the basic premise of current emergency planning regulations (i.e., likelihood of core damage) will be reduced to the extent that special emergency response planning beyond the exclusion area boundary may not be needed. How this very significant outcome can be effected through a highly risk-informed licensing is discussed. (author)

The intelligent customer: considerations around build-own-operate business and licensing models for small modular reactors in Canada

International Nuclear Information System (INIS)

Jones, K.

2014-01-01

An organization planning a proposal for a build-own-operate business model needs to address expanded licensee responsibilities under this model, associated regulatory impacts and how this affects their role as an 'intelligent customer'. This is particularly important for cases where builder-owner-operators plan to manufacture factory-fuelled designs and ship them to a site for installation and operation. The primary responsibility for safe conduct of licensed activities rests with the licensee. A build-own-operate model expands the scope of licensed activities to include design, manufacturing, transport, construction, and operation. The licensee must be able to demonstrate they are qualified to conduct all licensed activities including sufficient competent resources within the licensee's organization to oversee('Intelligent Customer') any work it commissions externally and the subsequent flow down through of the supply chain. This paper examines aspects that organizations need to assess the suitability of approaches that it may take to maintain in-house expertise for the control and oversight of licensed activities at all times. It considers the approach to identification of: core capabilities the licensee would need to understand its safety case under a build-own-operate model to manage licensed activities in accordance with requirements under the Nuclear Safety and Control Acta licensee's 'intelligent customer' capabilities in particular around understanding, specifying, overseeing and accepting work undertaken on its behalf by contractors. While this paper is focused on small modular reactors, being an intelligent customer applies to large commercial or research reactors equally; the size of reactor is immaterial.

The intelligent customer: considerations around build-own-operate business and licensing models for small modular reactors in Canada

Energy Technology Data Exchange (ETDEWEB)

Jones, K., E-mail: kenneth.jones@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

2014-07-01

An organization planning a proposal for a build-own-operate business model needs to address expanded licensee responsibilities under this model, associated regulatory impacts and how this affects their role as an 'intelligent customer'. This is particularly important for cases where builder-owner-operators plan to manufacture factory-fuelled designs and ship them to a site for installation and operation. The primary responsibility for safe conduct of licensed activities rests with the licensee. A build-own-operate model expands the scope of licensed activities to include design, manufacturing, transport, construction, and operation. The licensee must be able to demonstrate they are qualified to conduct all licensed activities including sufficient competent resources within the licensee's organization to oversee('Intelligent Customer') any work it commissions externally and the subsequent flow down through of the supply chain. This paper examines aspects that organizations need to assess the suitability of approaches that it may take to maintain in-house expertise for the control and oversight of licensed activities at all times. It considers the approach to identification of: core capabilities the licensee would need to understand its safety case under a build-own-operate model to manage licensed activities in accordance with requirements under the Nuclear Safety and Control Acta licensee's 'intelligent customer' capabilities in particular around understanding, specifying, overseeing and accepting work undertaken on its behalf by contractors. While this paper is focused on small modular reactors, being an intelligent customer applies to large commercial or research reactors equally; the size of reactor is immaterial.

Licensing decisions and safety research related to LMFBR accidents

International Nuclear Information System (INIS)

Denise, R.P.; Speis, T.P.; Kelber, C.N.; Curtis, R.T.

1977-01-01

The licensing approach which ensures adequate protection of the public health and safety against serious accidents is described. This paper describes the role of core melt and core disruptive accidents in the design, safety research, and licensing processes, using the Clinch River Breeder Reactor (CRBR) as a focal point. Major design attention is placed on the prevention of these accidents so that the probability of core melt accidents is reduced to a sufficiently low level that they are not treated as design basis accidents. Additional requirements are placed upon the design to further reduce residual risk. This licensing process is supported by a confirmatory research program designed to provide an independent basis for licensing judgements. It has as a goal the resolution of generic safety issues prior to the establishment of a commercial LMFBR industry. The program includes accident analysis, experiments in materials interactions, aerosol transport and system integrity and planning for new safety test facilities. The problems are approached in a multi-disciplinary functional manner that identifies key safety issues and centralizes efforts to resolve them. The near term objectives of the program support the licensing of the Clinch River Breeder Reactor (CRBR) and the proposed Prototype Large Breeder Reactor (PLBR). The long term objectives of the program support the licensing of commercial LMFBRs during the late 1980's and beyond. This safety research is designed to provide an independent basis for the licensing judgements which must be made by the Nuclear Regulatory Commission

Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

International Nuclear Information System (INIS)

1977-06-01

The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant

Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

International Nuclear Information System (INIS)

1989-04-01

This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

International Nuclear Information System (INIS)

Bacher, P.

1997-01-01

A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

Scaling analysis in bepu licensing of LWR

Energy Technology Data Exchange (ETDEWEB)

D' auria, Francesco; Lanfredini, Marco; Muellner, Nikolaus [University of Pisa, Pisa (Italy)

2012-08-15

'Scaling' plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a 'roadmap' to scaling. Key elements are the 'scaling-pyramid', related 'scaling bridges' and a logical path across scaling achievements (which constitute the 'scaling puzzle'). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the 'key-to-scaling', in the licensing process of a nuclear power plant. The proposed 'road map to scaling' aims at solving the 'scaling puzzle', by introducing a unified approach to the problem.

Sodium Fast Reactor Safety and Licensing Research Plan

International Nuclear Information System (INIS)

Denman, M.; Lachance, J.; Sofu, T.; Bari, R.; Flanagon, G.; Wigeland, R.

2015-01-01

This paper summarizes potential research priorities for the US Department of Energy (DOE) with the intent of improving the licensability of the sodium cooled fast reactor (SFR). In support of this project, five panels were tasked with identifying potential safety related gaps in the available information, data and models needed to support the licensing of an SFR. The areas examined were sodium technology; accident sequences and initiators; source term characterization, codes and methods; and fuels and materials. It is the intent of this paper to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the applied technology access control designation from old documents. The second cross-cutting gap is the need for a robust knowledge management and preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with applied technology and knowledge management. (author)

Safety-licensing assessment of NASAP reactor concepts and fuel cycle facilities

International Nuclear Information System (INIS)

Lipinski, W.C.; Prohammer, F.G.; van Erp, J.B.; Seefeldt, W.B.

1978-06-01

Assessments are presented of the safety/licensability of reactor concepts based on information supplied by the Nonproliferation Alternative Systems Assessment Program (NASAP) characterization contractors in their updated responses to the data package for NASAP Rolling Report II. The assessment of the LMFBR includes information from a characterization contractor on alternate fuel cycles but does not include information provided by a characterization contractor on plant-related safety issues. The information provided by the characterization contractors was supplemented by assessments provided by the U. S. Nuclear Regulatory Commission

ACR: Licensing and design readiness

International Nuclear Information System (INIS)

Alizadeh, A.

2009-01-01

Full text The Canadian nuclear technology has a long history dating back to the 1940s. In this regard, Canada is in a unique situation, shared only by a very few countries, where original nuclear power technology has been invented and further developed. Canadian Nuclear Safety Commission (CNSC), then called AECB, was established in 1946. CNSC focuses on nuclear security, nuclear safety, establishing health and safety regulations, and has also played an instrumental role in the formation of the IAEA. CNSC has provided assistance to the establishment of regulatory authorities in AECL's client countries such as Korea, Argentina, China and Romania. AECL has developed the Gen III+ ACR 1000 as evolutionary advancement of the current CANDU 6 reactor. ACR-1000 has evolved from AECL's in depth experience with CANDU systems, components, and materials, as well as the feedback received from owners and operators of CANDU plants. The ACR-1000 design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. It also features major improvements in economics, inherent safety characteristics, and performance. ACR-1000 has completed its Basic Engineering, has advanced in the licensing process in Canada, and is ready for deployment in Canadian and world markets. EC6 is an evolution of CANDU 6 and is a Gen III natural uranium fuelled reactor. Its medium size and potential for fuel localization and advanced fuel cycles is an optimal strategic solution in many markets.AECL's reactor products are shown to be compliant with a variety of licensing and regulatory requirements. These include the new CNSC DRD-337, IAEA NS-R1, and EUR. This allows the countries interested in CANDU reactor products to be confident of its licensing in their own regulatory regimes.

The Reactor Analysis Support Package (RASP): Volume 10, Guidelines for developing a reload licensing capability:Final report

International Nuclear Information System (INIS)

Diamond, D.J.

1988-08-01

The EPRI Reactor Analysis Support Package (RASP) consists of computer codes and documentation for calculating core performance and plant transients. This report was written to help a utility to use these tools properly for reload design and safety evaluations. The emphasis is on the steps that the utility can take to develop a methodology that is approved by the Nuclear Regulatory Commission for reload licensing submittals. The report treats both the planning and the implementation of this type of calculational capability. With regard to planning, there is discussion defining objectives, resource requirements, organization and scheduling. In order to help the engineering staff implement the plan there is discussion of the development of a methodology for event analysis, qualification of the methods, and the writing of design control procedures and topical reports. The experience of utilities, and especially of GPU Nuclear Corporation (GPUN), in developing a reload licensing capability is cited throughout the report and extracts from GPUN design control procedures are included in the appendices. 16 refs., 23 figs., 9 tabs

Operator licensing examiner standards

International Nuclear Information System (INIS)

1994-06-01

The Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining licensees and applicants for reactor operator and senior reactor operator licenses at power reactor facilities pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). The Examiner Standards are intended to assist NRC examiners and facility licensees to better understand the initial and requalification examination processes and to ensure the equitable and consistent administration of examinations to all applicants. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator licensing policy changes. Revision 7 was published in January 1993 and became effective in August 1993. Supplement 1 is being issued primarily to implement administrative changes to the requalification examination program resulting from the amendment to 10 CFR 55 that eliminated the requirement for every licensed operator to pass an NRC-conducted requalification examination as a condition for license renewal. The supplement does not substantially alter either the initial or requalification examination processes and will become effective 30 days after its publication is noticed in the Federal Register. The corporate notification letters issued after the effective date will provide facility licensees with at least 90 days notice that the examinations will be administered in accordance with the revised procedures

Safety philosophy and licensing practice in different member states of IAEA: Spain

International Nuclear Information System (INIS)

Alonso, A.

1981-01-01

With the aim of presenting the licensing experience in Spain, the countries with nuclear activities are divided into three main groups: exporters, qualified importers and importers, being Spain in the second group. The licensing problems, and therefore the licensing experience, are different in the different groups of countries. Moreover, the experience in Spain is enriched by the following facts: an early start, a substantial program and a diversity of types, pressurized and boiling water reactors, and suppliers, American and German plants. Reference is made to the basic legal documents governing licensing in Spain, together with the difficulties in adopting and applying detailed regulations. Within this framework, the licensing experience in Spain is described with reference to the reference plant concept, later enlarged to include the reference-site reference plant concept, ending in the most practical approach of the reference problem concept. Finally specific licensing problems are introduced with reference to the Spanish operating nuclear power plants, the ones under commissioning and in an advanced state of construction and the ones just staring construction. (orig./RW)

The pre-licensing of a multi-purpose hybrid research reactor for high-tech applications 'MYRRHA'

International Nuclear Information System (INIS)

Hakimi, N.; Dams, C.; Wertelaers, A.; Nys, V.; Schrauben, M.; Dresselaers, R.

2013-01-01

The Belgian Nuclear Research Centre in Mol has been working for several years on the design of a multi-purpose flexible irradiation facility in order to replace the ageing BR2, a multipurpose materials testing reactor (MTR), in operation since 1962. MYRRHA, a flexible fast spectrum research reactor is conceived as an accelerator driven system (ADS), able to operate in sub-critical and critical modes. It contains a proton accelerator of 600 MeV, a spallation target and a multiplying medium with MOX fuel, cooled by liquid lead-bismuth (Pb-Bi). Since February 2011, the Belgian Nuclear Research Centre has engaged in a 'pre-licensing' process with the regulatory authority for an estimated period up to mid 2014. The paper presents on the one hand the objectives of the pre-licensing phase as well as its implementation process and on the other hand, 2 implementing instruments which have been developed by the regulatory authority providing guidance to the designer of MYRRHA in order to meet the pre-licensing phase objectives. The first instrument is a strategic note for the design and operation of MYRRHA where as the second instrument is a guidance document for the format and content of a design options and provisions file (DOPF). Both instruments have been developed taking into account that MYRRHA is an irradiation facility using a Generation IV nuclear power system's type technology (liquid metal cooled fast neutron reactor). The strategic note overview aims to cover the safety approach as well as the security requirements and safeguards obligation applicable to MYRRHA. In particular, in the strategic note, a specific attention has been paid in order to ensure that a safety, security and safeguards integrated approach will drive the development of the MYRRHA design. The safety approach focuses on the safety goals and the minimum safety objectives set by the regulatory authority for this innovative design. The DOPF overview presents its objectives and structure resuming

Reactors licensing: proposal of an integrated quality and environment regulatory structure for nuclear research reactors in Brazil; Licenciamento de reatores: proposta de uma estrutura regulatoria integrada com abordagem em qualidade e meio ambiente para reatores de pesquisa no Brasil

Energy Technology Data Exchange (ETDEWEB)

Serra, Reynaldo Cavalcanti

2014-07-01

A new integrated regulatory structure based on quality and integrated issues has been proposed to be implemented on the licensing process of nuclear research reactors in Brazil. The study starts with a literature review about the licensing process in several countries, all of them members of the International Atomic Energy Agency. After this phase it is performed a comparative study with the Brazilian licensing process to identify good practices (positive aspects), the gaps on it and to propose an approach of an integrated quality and environmental management system, in order to contribute with a new licensing process scheme in Brazil. The literature review considered the following research nuclear reactors: Jules-Horowitz and OSIRIS (France), Hanaro (Korea), Maples 1 and 2 (Canada), OPAL (Australia), Pallas (Holand), ETRR-2 (Egypt) and IEA-R1 (Brazil). The current nuclear research reactors licensing process in Brazil is conducted by two regulatory bodies: the Brazilian National Nuclear Energy Commission (CNEN) and the Brazilian Institute of Environment and Renewable Natural Resources (IBAMA). CNEN is responsible by nuclear issues, while IBAMA by environmental one. To support the study it was applied a questionnaire and interviews based on the current regulatory structure to four nuclear research reactors in Brazil. Nowadays, the nuclear research reactor’s licensing process, in Brazil, has six phases and the environmental licensing process has three phases. A correlation study among these phases leads to a proposal of a new quality and environmental integrated licensing structure with four harmonized phases, hence reducing potential delays in this process. (author)

Dry spent fuel storage licensing

International Nuclear Information System (INIS)

Sturz, F.C.

1995-01-01

In the US, at-reactor-site dry spent fuel storage in independent spent fuel storage installations (ISFSI) has become the principal option for utilities needing storage capacity outside of the reactor spent fuel pools. Delays in the geologic repository operational date at or beyond 2010, and the increasing uncertainty of the US Department of Energy's (DOE) being able to site and license a Monitored Retrievable Storage (MRS) facility by 1998 make at-reactor-site dry storage of spent nuclear fuel increasingly desirable to utilities and DOE to meet the need for additional spent fuel storage capacity until disposal, in a repository, is available. The past year has been another busy year for dry spent fuel storage licensing. The licensing staff has been reviewing 7 applications and 12 amendment requests, as well as participating in inspection-related activities. The authors have licensed, on a site-specific basis, a variety of dry technologies (cask, module, and vault). By using certified designs, site-specific licensing is no longer required. Another new cask has been certified. They have received one new application for cask certification and two amendments to a certified cask design. As they stand on the brink of receiving multiple applications from DOE for the MPC, they are preparing to meet the needs of this national program. With the range of technical and licensing options available to utilities, the authors believe that utilities can meet their need for additional spent fuel storage capacity for essentially all reactor sites through the next decade

Technology neutral licensing requirements: have we been successful?

International Nuclear Information System (INIS)

Jamieson, T.

2009-01-01

The Canadian Nuclear Safety Commission (CNSC) is Canada's nuclear regulator. In preparation for the licensing of the next generation of nuclear reactors in Canada, the CNSC has made major changes in its processes and procedures associated with all aspects of the licensing life cycle. These include: development of key regulatory documents outlining the CNSC expectations for the design and siting of new reactors; creation of applicant guidelines for completing licensing applications; creation of a comprehensive set of review guides, specifying how topics in applications are to be reviewed; performing pre-project design reviews for proposed new reactor technologies; and streamlining processes wherever possible, such as performing the environmental assessment and license to prepare site activities in parallel. The predicted workload associated with the nuclear renaissance in Canada is real. The CNSC is currently performing: 4 pre-project design reviews (for AECL's ACR-1000 and EC-6, Areva's EPR and Westinghouse's AP-1000), has received applications for 4 new builds (at Darlington, Bruce, Nanticoke and in Alberta) and is in the midst of performing the combined environmental assessment and license to prepare site phases for these sites. This is in addition to the work associated with: the licensing of new uranium mines; oversight of the existing reactor fleet; refurbishments at Bruce/Gentilly/Point Lepreau; license renewal for the NRU reactor; and oversight of our general client base of over 5000 licensees. In this presentation, experience to date with these new processes and procedures will be discussed, including: status of the current new build projects; anticipated new activities; lessons learned, especially the technology neutral approach; application to our other licensed activities, including non-power reactors; and work underway to further develop CNSC methods. Qualifications Needed to Design, Construct and Manufacture Nuclear Systems and Equipment, C. Voutsinos

U.S. regulatory requirements for nuclear plant license renewal: The B and W Owners Group License Renewal Program

International Nuclear Information System (INIS)

Staudinger, Deborah K.

2004-01-01

This paper discusses the current U.S. Regulatory Requirements for License Renewal and describes the Babcock and Wilcox Owners Group (B and WOG) Generic License Renewal Program (GLRP). The B and W owners, recognizing the need to obtain the maximum life for their nuclear generating units, embarked on a program to renew the licenses of the seven reactors in accordance with the requirements of the Atomic Energy Act of 1954 and further defined by Title 10 of the Code of Federal Regulation Part 54 (10 CFR 54). These reactors, owned by five separate utilities, are Pressurized Water Reactors (PWR) ranging in net rated capacity from approximately 800 to 900 MW. The plants, predominately constructed in the 70s, have USNRC Operating Licenses that expire between 2013 to 2017. (author)

Recent U.S. reactor operating experience

International Nuclear Information System (INIS)

Stello, V. Jr.

1977-01-01

A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

The United States nuclear regulatory commission license renewal process

International Nuclear Information System (INIS)

Holian, B.E.

2009-01-01

The United States (U.S.) Nuclear Regulatory Commission (NRC) license renewal process establishes the technical and administrative requirements for the renewal of operating power plant licenses. Reactor ope-rating licenses were originally issued for 40 years and are allowed to be renewed. The review process for license renewal applications (L.R.A.) provides continued assurance that the level of safety provided by an applicant's current licensing basis is maintained for the period of extended operation. The license renewal review focuses on passive, long-lived structures and components of the plant that are subject to the effects of aging. The applicant must demonstrate that programs are in place to manage those aging effects. The review also verifies that analyses based on the current operating term have been evaluated and shown to be valid for the period of extended operation. The NRC has renewed the licenses for 52 reactors at 30 plant sites. Each applicant requested, and was granted, an extension of 20 years. Applications to renew the licenses of 20 additional reactors at 13 plant sites are under review. As license renewal is voluntary, the decision to seek license renewal and the timing of the application is made by the licensee. However, the NRC expects that, over time, essentially all U.S. operating reactors will request license renewal. In 2009, the U.S. has 4 plants that enter their 41. year of ope-ration. The U.S. Nuclear Industry has expressed interest in 'life beyond 60', that is, requesting approval of a second renewal period. U.S. regulations allow for subsequent license renewals. The NRC is working with the U.S. Department of Energy (DOE) on research related to light water reactor sustainability. (author)

Operational experience with Dragon reactor experiment of relevance to commercial reactors

International Nuclear Information System (INIS)

Capp, P.D.; Simon, R.A.

1976-01-01

An important part of the experience gained during the first ten years of successful power operation of the Dragon Reactor is relevant to the design and operation of future High Temperature Reactors (HTRs). The aspects presented in this paper have been chosen as being particularly applicable to larger HTR systems. Core performance under a variety of conditions is surveyed with particular emphasis on a technique developed for the identification and location of unpurged releasing fuel and the presence of activation and fission products in the core area. The lessons learned during the reflector block replacement are presented. Operating experience with the primary circuit identifies the lack of mixing of gas streams within the hot plenum and the problems of gas streaming in ducts. Helium leakage from the circuit is often greater than the optimum 0.1%/d. Virtually all the leakage problems are associated with the small bore instrument pipework essential for the many experiments associated with the Dragon Reactor Experiment (DRE). Primary circuit maintenance work confirms the generally clean state of the DRE circuit but identifies 137 Cs and 110 Agsup(m) as possible hazards if fuel emitting these isotopes is irradiated. (author)

Licensing aspects in the verification of the SNR 300 design concept against hypothetical accidents

International Nuclear Information System (INIS)

Kugler, E.; Wiesner, S.

1976-01-01

The German prototype of a fast breeder reactor, the SNR 300, is being built near Kalkar on the Lower Rhine. It is a loop-type fast sodium-cooled reactor, designed and constructed by Interatom, Bensberg. Experiences gained from the first phase of construction are described. The report is restricted to the aspects of the SNR 300 design against a core disruptive accident (CDA) and its consequences and to the difficulties having arisen in the verification of the design concept so far. Some examples of the detailed design are described and discussed from the licensing authority's point of view showing that the difficulties have been typical for a prototype reactor subjected to a regular licensing procedure

Impact of 2D/3D-project on LOCA-licensing analysis and reactor safety of PWRs

International Nuclear Information System (INIS)

Winkler, F.; Krebs, W.D.

1989-01-01

In the past LOCA-licensing analysis has included large conservatisms to compensate for the lack of detailed two phase flow and full scale experimental data. The 2D/3D-project was established to improve the data base in order to minimize the conservatisms required. The significant results and findings of the full scale Upper Plenum Test Facility (UPTF) and from the electrically heated Slab Core Test Facility (SCTF) were particularly useful for understanding the multidimensional phenomena in the primary system and in the core of a PWR. UPTF results were used to verify the TRAC-PF1 analysis of a PWR with combined ECC-Injection during the reflood phase of a large break-LOCA. Comparison of these results with results from classic licensing calculations quantifies the large safety margin in earlier licensing procedures and in reactor systems. (orig.)

License renewal

International Nuclear Information System (INIS)

Newberry, S.

1993-01-01

This article gives an overview of the process of license renewal for nuclear power plants. It explains what is meant by license renewal, the significance of license renewal, and goes over key elements involved in the process of license renewal. Those key elements are NRC requirements embodied in 10 CFR Part 54 (Reactor Safety) and 10 CFR Part 51 (Environmental Issues). In addition Industry Reports must be developed and reviewed. License renewal is essentially the process of applying for a 20 year extension to the original 40 year operating license granted for the plant. This is a very long term process, which involves a lot of preparation, and compliance with regulatory rules and guidelines. In general it is a process which is expected to begin when plants reach an operating lifetime of 20 years. It has provisions for allowing the public to become involved in the review process

Experience in the recruitment, organization and training of operations and maintenance personnel for the Malaysian research reactor

International Nuclear Information System (INIS)

Jamal Khair Ibrahim.

1983-01-01

The TRIGA Reactor located at the Tun Ismail Atomic Research Centre (PUSPATI) Complex is owned and operated by the Nuclear Energy Unit of the Prime Minster's Department. The operations and maintenance personnel are part and parcel of the national civil service organization. As such, the requirement and remuneration of these personnel are handled by a central federal government personnel management agency in common with personnel from other federal government agencies. In addition, the reactor is the first and only one in Malaysia, a developing country, which is the process of committing herself towards a nuclear power programme. These factors coupled with the absence of an independent reactor operator licensing agency posed unique problems in the recruitment, organization, training and licensing of operations personnel for the facility. The paper discusses these factors and their bearing on the recruitment, training, licensing and career development prospects of the PUSPATI TRIGA Reactor operators. (author)

Conceptual framework for using 'Best Estimate plus Uncertainty' as a basis for licensing activities for fuels developed for an advanced reactor

International Nuclear Information System (INIS)

McClure, P.; Unal, C.; Boyack, B.

2010-01-01

Closing the fuel cycle is one of the major technical challenges to expanding the use of nuclear energy to meet the world's need for benign, environmentally safe electrical power. 'Closing the fuel cycle ' means getting the maximum amount of energy possible out of uranium fuel while minimizing the amount of high-level waste that must be stored. The U.S. Dept. of Energy's Fuel Cycle Research and Development (FCRD) program is investigating the recycling of transuranic isotopes contained in spent nuclear fuel. Recycling minimizes the amount of high-level waste that would require storage in repositories. Developing new fuels and the advanced reactors that burn them is a long process typically spanning two decades from concept to final licensing. A unique challenge to meeting the FCRD objectives in this area is the fact that the experimental database is incomplete. Thus, using a traditional, heavily empirical approach to develop and qualify fuels for an advanced reactor plant will be very challenging. To address this concern, FCRD has launched an advanced modeling and simulation (M and S) approach to revolutionize fuel development and advanced reactor design. This new approach depends on transferring recent advances in the computational sciences and computer technologies into the development of these program elements. The licensing process that historically has been used by the U.S. Nuclear Regulatory Commission (NRC) for fuels qualification is based on using a large body of experimental work to qualify and license a new fuel. If an M and S approach with more directed experimentation is to be considered as an alternative approach for licensing, a framework needs to be developed early in the process. Using M and S with limited experiments as a basis for demonstrating that a design can meet NRC requirements is not new and has precedence in the NRC. The method is generically referred to as a 'Best Estimate plus Uncertainty' (BE+U) approach because the goal of the

Sodium fast reactor safety and licensing research plan. Volume I.

Energy Technology Data Exchange (ETDEWEB)

Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

2012-05-01

This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

Sodium fast reactor safety and licensing research plan - Volume I

International Nuclear Information System (INIS)

Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

2012-01-01

This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

The Dragon reactor experiment

International Nuclear Information System (INIS)

Anon.

1976-01-01

The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

Safety Evaluation Report related to the renewal of the operating license for the research reactor at Pennsylvania State University

International Nuclear Information System (INIS)

1986-01-01

This Safety Evaluation Report for the application filed by the Pennsylvania State University for a renewal of Operating License R-2 to continue to operate the Pennsylvania State University Breazeale Reactor (PSBR) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the campus in University Park, Pennsylvania. On the basis of its technical review, the staff concludes that the reactor facility can continue to be operated by the university without endangering the health and safety of the public or the environment

Regulations concerning licensing of nuclear reactor facilities and other nuclear installations, Decree No 7/9141, 6 January 1975

International Nuclear Information System (INIS)

1975-01-01

This Decree lays down the licensing system for nuclear installations in Turkey and also sets up a Nuclear Safety Committee whose duty is to ensure that the requirements of this Decree are met. The Committee is made up of members of the Atomic Energy Commission specialized in reactors, nuclear safety, health physics, reactor physics as well as two experts respectively appointed by the Ministry of Health and Social Welfare and the Ministry of Energy and National Resources. (NEA) [fr

The Licensing of New Nuclear Power Plants in Europe

International Nuclear Information System (INIS)

Raetzke, C.

2008-01-01

After an introduction dealing with the nuclear Renaissance in Europe and the specific situation of Germany and of Italy, the article focuses on the question of licensing processes for new reactors. New nuclear power plant projects involve a substantial investment and electric utilities will only take this decision if the licensing and regulatory risk can be adequately managed. Licensing processes should be predictable and efficient in order to give sufficient assurance to applicants. The article discusses best practice in licensing by giving some examples of suitable licensing processes of other countries. It also highlights international initiatives aimed at harmonizing safety requirements for new reactors and a multinational cooperation in reactor design review. These issues should be carefully considered by any country wanting to get new nuclear started. [it

Current status of the PBMR licensing project

International Nuclear Information System (INIS)

Mysen, A.; Clapisson, G.A.; Metcalf, P.E.

2000-01-01

The CNS is currently reviewing the PBMR conceptual design from a licensibility point of view. The PBMR concept is based on a High Temperature Gas Cooled Reactor - pebble bed reactor type. It is anticipated that the PBMR design will rely on inherent safety characteristics to contain fission products within fuel over the full range of design basis events. This feature combined with the high temperature integrity of the fuel and structural graphite, allows the safe use of a high coolant temperature, which allows consideration of the future development of this reactor for non-electrical applications of nuclear heat for industrial use. The CNS licensing approach requires that the licensing and design basis of the plant should respect prevailing international norms and practices and that a quantitative risk assessment should demonstrate compliance with the CNS fundamental safety standards. The first stage of the licensing process is now ongoing; this is a pre-application phase, which will result in a statement on licensibility being issued. Identification of the specific documentation requirements and information needed is required across every step of the licensing process. Top level regulatory requirements have been established for the PBMR. They include the CNS fundamental safety standard and basic licensing criteria, which describes requirements on licensees of nuclear installations regarding risk assessment and compliance with the safety criteria and define classification of licensing basis events. (author)

Shutdown requirements in licensing of research reactors and consequences for operation

International Nuclear Information System (INIS)

Roegler, H.-J.; Stein, J.

1983-01-01

Most countries start their nuclear life with the setup of a research reactor. Most countries have at this time no licensing authority established which can act on existing laws, rules and regulations. Thus, most countries have during this starting phase to rely on the regulations fixed by international institutions such as the IAEA (or by national institutions as the former NRC in US, which happens to take place more often from reasons below). This implies a specific responsibility for international regulations since they have to be very clear and unique. Moreover they have to be applicable to all plants they are set forth for and they have to give no unavoidable restrictions on the use of the plant. Additionally some recheck of the rules based on national rules in forth will be of help. The enrichment reduction for research reactors gave frequent cause to look at the existing IAEA-publications for research reactors. They are laid down in the safety series no. 35 edited 1971. I will only discuss here specific parts of these considerations or requirements. Since the authors' personal impressions are that the published considerations are neither that clear nor that less restrictive as necessary the subsequent discussion is carried out. One of the confusing points of the existing IAEA-regulations is the lengthy discussion of reactivity considerations for critical assemblies first, no similar separate discussion for research reactors and a second discussion titled 'control-rod considerations' under the main title 'considerations applying to critical assemblies and research reactor'. This discussion will be restricted to the latter since the critical assemblies with their timely Constance of maximum excess reactivity do not present any problem for all shut-down conditions

Recent developments in Canadian nuclear power plant licensing practices

International Nuclear Information System (INIS)

Marchildon, P.

1985-06-01

This paper examines the dominant factors which have influenced the safety evaluation and licensing process of current 600 MW reactors. It describes possible modifications to the process which are being considered for the licensing of a second 600 MW reactor at Point Lepreau. The key element is a firm licensing agreement covering the entire licensing cycle, to be established between the proponent and the AECB before a construction licence is issued. Progress accomplished to date in reaching such an agreement is described

Challenges of SMR licensing practices

Energy Technology Data Exchange (ETDEWEB)

Soderholm, K., E-mail: kristiina.soderholm@fortum.com [Fortum Power, Espoo (Finland)

2012-12-15

This paper aims to increase the understanding of high level Nuclear Power Plant (NPP) licensing processes in Finland, France, the UK, Canada and the USA. These countries have been selected for this study because of their different licensing processes and recent actions in new NPP construction. After discussing their similarities and differences, suitable features for Small Modular Reactor licensing can be emphasized and suggested. Some of the studied licensing processes have elements that are already quite well suited for application to SMRs, but all of these different national processes can benefit from studying and implementing lessons learned from SMR specific licensing needs. The main SMR features to take into account in licensing are standardization of the design, modularity, mass production and serial construction. Modularity can be divided into two different categories: the first category is simply a single unit facility constructed of independently engineered modules (e.g., construction process for Westinghouse AP-1000 NPP) and the second is a facility structure composed of many reactor modules where modules are manufactured in factories and installed into the facility as needed (e.g., NuScale Power SMR design). Short construction schedules will not be fully benefited from if the long licensing process prolongs the commissioning and approach to full-power operation. The focus area of this study is to better understand the possibility of SMR deployment in small nuclear countries, such as Finland, which currently has four operating NPPs. The licensing process needs to be simple and clear to make SMR deployment feasible from an economical point of view. This paper uses public information and interviews with experts to establish the overview of the different licensing processes and their main steps. A high-level comparison of the licensing steps has been carried out. Certain aspects of the aviation industry licensing process have also been studied and certain

Application of up-front licensing

Energy Technology Data Exchange (ETDEWEB)

Grant, S D [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada); Snell, V G [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

AECL has been pioneering `up-front` licensing of new reactor designs. The CANDU 3 design has been formally reviewed by AECB staff for a number of years. The CANDU 9 design has just started the up-front licensing process. The process gives designers, regulators and potential customers early confidence in the licensability of future plants. (author). 4 refs., 2 tabs.

Application of up-front licensing

International Nuclear Information System (INIS)

Grant, S.D.; Snell, V.G.

1995-01-01

AECL has been pioneering 'up-front' licensing of new reactor designs. The CANDU 3 design has been formally reviewed by AECB staff for a number of years. The CANDU 9 design has just started the up-front licensing process. The process gives designers, regulators and potential customers early confidence in the licensability of future plants. (author). 4 refs., 2 tabs

In-pile experiments and test facilities proposed for fast reactor safety

International Nuclear Information System (INIS)

Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

1976-01-01

The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

ABWR certification work brings US licensing stability nearer

International Nuclear Information System (INIS)

Wilkins, D.R.; Quirk, J.F.

1991-01-01

The Advanced Boiling Water Reactor (ABWR) is now approaching Final Design Approval by the US Nuclear Regulatory Commission (NRC) and will then proceed on to the certification phase of the NRC's new standard plant licensing process. Successful completion of this will usher in a new era of standardization and reactor licensing stability in the US. (author)

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

Energy Technology Data Exchange (ETDEWEB)

Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

2017-07-01

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

Licensed operating reactors. Status summary report, data as of October 31, 1985. Volume 9, No. 11

International Nuclear Information System (INIS)

1985-12-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the errata page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States. The percentage computations, Items 20 through 24 in Section 2, the vendor capacity factors on page 1-7, and actual vs potential energy production on page 1-2 are computed using actual data for the period of consideration. The percentages listed in power generation on page 1-2 are computed as an arithmetic average. The factors for the life-span of each unit (the ''Cumulative'' column) are reported by the utility and are not entirely re-computed by NRC. Utility power production data is checked for consistency with previously submitted statistics

Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

International Nuclear Information System (INIS)

1987-05-01

Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions

U.S. licensing process and ABWR certification

International Nuclear Information System (INIS)

Quirk, J.F.; Williams, W.A.

1996-01-01

Part 50 of Title 10 of the Code of Federal Regulation (CFR) establishes a two-step licensing process by which the U.S. Nuclear Regulatory Committee (NRC) authorizes nuclear reactor plant construction through issuance of a construction permit and authorizes operation by issuance of an operating license. At each stage, the NRC Staff conducts technical reviews and there is potential for public hearings. In 1989, the NRC issued a new, simplified licensing process: Part 52. The purpose of the Part 52 licensing process is to provide a regulatory framework that brings about earlier resolution of licensing issues. Because issues are not resolved early in the Part 50 licensing process, approval of an operating license is not assured until after a significant investment has been made in the plant. Part 52 increases the stability and certainty of the licensing process by providing for the early resolution of safety and environmental issues. The Part 52 licensing process features (1) early site permits, (2) design certification, and (3) combined construction permit and operating licenses. As part of the U.S. Advanced Light Water Reactor (ALWR) Program to revitalize the nuclear option through the integration of government/utility/industry efforts, GE undertook the role of applying for certification for its latest product line, the Advanced Boiling Water Reactor (ABWR), under the U.S. ABWR certification program. The ABWR design is an essentially complete plant. Initial application for design certification was in 1987 under Part 50. GE reapplied in late 1991 under the newly promulgated Part 52. Following seven years of intensive interactions with the NRC and ACRS, GE was awarded the first Final Design Approval (FDA) under Part 52. The Commission initiated rulemaking by publishing the proposed ABWR Certification Rule in the Federal Register in early 1995. Certification is anticipated mid-1996. (J.P.N.)

Study on reactor power transient characteristics (reactor training experiments). Control rod reactivity calibration by positive period method and other experiment

International Nuclear Information System (INIS)

Ozaki, Yoshihiko; Sunagawa, Takeyoshi

2014-01-01

In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity calibration experiments using positive Period method were carried out for shim safety rod and regulating rod, respectively. The results were obtained as reasonable values in comparison with the nominal value of the UTR-KINKI. The measurement of reactor power change after reactor scram was performed, and the presence of the delayed neutron precursor was confirmed by calculating the half-life. The spatial dose rate measurement experiment of neutrons and γ-rays in the reactor room in a reactor power 1W operating conditions were also performed. (author)

Safety of next generation power reactors

International Nuclear Information System (INIS)

Anon.

1988-01-01

This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

Licensing of ''grandfather's'' facilities: Ukrainian experience

International Nuclear Information System (INIS)

Mikolaitchouk, H.; Bogdan, L.; Steinberg, N.

1995-01-01

In the former USSR, unlike most countries, radioactive waste management activities including waste disposal needed no license. But after the USSR breakdown the Ukrainian Parliament -- Verkhovna Rada -- invoked the revised Law on Business activities. According to Article 4 of the Law, in order to treat or to dispose radioactive waste every enterprise has to get a special permission or license. In compliance with the Law, the Cabinet of Ministers by its Ordinance of January 13, 1993, authorized the Ukrainian State Committee for Nuclear and Radiation Safety (UkrSCNRS) to issue special permissions or licenses for waste treatment and disposal. And that requirement was valid not only for future activities but also for existing facilities in operation. Taking into account the undergoing legislative process, SCNRS began to develop its licensing process without waiting for the special nuclear laws to be passed. On the basis of the legislation already in effect, first of all the Law on Enterprises (full responsibility of enterprises for their activities) and Law on Business activities (requirement to have a license for special types of activities), the newly formed national regulatory body had to identify all the enterprises that needed to be licensed, to establish relevant procedures, to develop related regulatory documents, to implement these procedures and documents at operating enterprises, and for each case to make a decision concerning feasibility of issuing a license, period of validity and license conditions

KNK II, Compact Sodium-Cooled Reactor in the Nuclear Research Center Karlsruhe

International Nuclear Information System (INIS)

1978-01-01

The report gives an overview of the project of the sodium-cooled fast reactor KNK II in the nuclear research center KfK in Karlsruhe. This test reactor was the preparatory stage of the prototype plant SNR 300 and had several goals: to train operating personal, to practice the licensing procedures in Germany, to get experience with the sodium technology and to serve as a test bed for fast breeder core components. The report contains contributions of KfK as the owner and project managing organization, of INTERATOM as the design and construction company and of the KBG as the plant operating organization. Experience with and results of relevant aspects of the project are tackled: project management, reactor core and component design, safety questions and licensing, plant design and test programs [de

Safety Evaluation Report related to the renewal of the operating license for the Westinghouse research reactor at Zion, Illinois (Docket No. 50-87)

International Nuclear Information System (INIS)

1984-09-01

This Safety Evaluation Report, for the application filed by the Westinghouse Electric Company, for renewal of operating license number R-119 to continue to operate the research reactor, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is operated by Westinghouse and is located in Zion, Illinois. The staff concludes that the reactor facility can continue to be operated by Westinghouse without endangering the health and safety of the public

Safety-evaluation report related to the renewal of the operating license for the Cornell University TRIGA Research Reactor. Docket No. 50-157

International Nuclear Information System (INIS)

1983-08-01

This Safety Evaluation Report for the application filed by the Cornell University for a renewal of Operating License R-80 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by Cornell University and is located on the Cornell campus in Ithaca, New York. The staff concludes that the TRIGA reactor facility can continue to be operated by Cornell without endangering the health and safety of the public

An example of regulating process for an advanced reactor: Creys-Malville

International Nuclear Information System (INIS)

Cravero, M.

1978-01-01

The general philosophy for the regulatory control of the Super-Phenix fast breeder reactor was the same as for other types of nuclear power plants. The licensing process for both conventional and nuclear aspects was also identical to that for other nuclear units. On the other hand, safety recommendations were especially prepared by the CEA's appropriate services on the basis of acquired experience. The paper analyses the development of the licensing process for the reactor and knowledge to be gained for future units of this type. (NEA) [fr

Federal supervisory authorities' power to issue directives in nuclear licensing procedures (Kalkar reactor)

International Nuclear Information System (INIS)

Anon.

1990-01-01

The Federal Constitutional Court (BVerfG) is concerned in its judgement with the legal instrument of Federal supervisory authorities' power to issue directives in nuclear licensing procedures (Kalkar reactor). Dealt with are questions concerning material and excercising competences, the Federal Government's right to issue directives and possible violation of Laender rights, as well as the legal position between Land and Federal Government, the necessity to issue clear directives, questions concerning the competence to issue directives and the Federal Government's duty to consider the overall interests (Federal Government and Laender interests), questions concerning constitutional state principles and the limits in the relations between Federal Government - Laender concerning legal competence. (RST) [de

Operation and maintenance experiences at the C.R.E. Casaccia TRIGA reactor

International Nuclear Information System (INIS)

Festinesi, A.

1988-01-01

The memoir explains TRIGA RC-1 plant activities from last European TRIGA Users' Conference till today. In particular, measures following reactor exercise license renewing (March 1987) are described. Finally, difficulties and measures about shielding tank's water funguses and spores contamination, are explained. (author)

Safety Evaluation Report related to the renewal of the operating license for the research reactor at Purdue University: Docket No. 50-182

International Nuclear Information System (INIS)

1988-04-01

This Safety Evaluation Report for the application filed by Purdue University for a renewal of Operating License R-87 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned by Purdue University and is located on the campus in West Lafayette, Indiana. On the basis of its technical review, the staff concludes that the reactor facility can continue to be operated by the university without endangering the health and safety of the public or the enviroment

Safety evaluation report related to the renewal of the operating license for the Washington State University TRIGA reactor. Docket No. 50-27

International Nuclear Information System (INIS)

1982-05-01

This Safety Evaluation Report for the application filed by the Washington State University (WSU) for a renewal of operating license number R-76 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Washington State University and is located on the WSU campus in Pullman, Whitman County, Washington. The staff concludes that the TRIGA reactor facility can continue to be operated by WSU without endangering the health and safety of the public

The changing shape of U.S. licensing philosophy

International Nuclear Information System (INIS)

Remick, F.J.

1992-01-01

The shape of U.S. nuclear licensing and regulatory philosophy and process has already changed. The new process requires NRC review and approval of the vendor designs before a prospective utility license applicant purchases the design and begins construction. The new philosophy has resulted from the lessons learned from extensive operating experience accumulated in the United States. New criteria established for judging reactor designs include the capability of future designs to be more tolerant of accidents beyond the traditional design basis events. Qualitative and quantitative goals have been chosen as a guide for allocating resources for regulation of the currently operating plants. The changing shape of nuclear licensing and regulatory philosophy is also a result of economic circumstances in the United States. All will have a better opportunity to take part in the process which is most likely to encourage further development of safe nuclear energy in the United States. (author)

PARR-2: reactor description and experiments

International Nuclear Information System (INIS)

Wyne, M.F.; Meghji, J.H.

1990-12-01

PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

Development regulation regarding with licensing of nuclear installation

International Nuclear Information System (INIS)

Bambang Riyono; Yudi Pramono; Dahlia Cakrawati Sinaga

2011-01-01

Provisions of Article 17 paragraph (3) of Law Number 10 Year 1997 on Nuclear cleary mandates for the establishment of government regulations (GR) on Nuclear licensing containing the requirements and procedure, both from the standpoint of their utilization and installation. To use has been rising GR No.29 Year 2008 on the Use of Ionizing Radiation Sources and Nuclear Materials, while for the installation has been published PP No.43 Year 2006 on Nuclear Reactor Licensing, and BAPETEN Chairman Decree No.3 Year 2006 on Non-reactor Nuclear Installation Licensing. Based on the background of the preparation of both the aforementioned are just regulate the reactor and utilization, not yet fully meet the mandate of Article 17 paragraph (3) of Law No.10 of 1997 on Nuclear, including other nuclear installations. For these reasons, it initiated the need for a separate regulation containing provisions concerning licensing of non-reactor nuclear installations. On the other side from the understanding the legal aspects and interpretations of the Law No.10 of 2004 on the Establishment Regulation Legislation, should be in single mandate of Article 17 paragraph (3) of Law No.10 of 1997 on Nuclear would only produce one of the requirements and procedure for the use or installation, or a maximum of two (2) GR related licensing the use and installation. This is encourages conducted the assessing or studies related to how possible it is according to the legal aspect is justified to combine in one Nuclear licensing regulations regarding both the use and installation, by looking at the complexity of installation and wide scope of utilization of nuclear energy in Indonesia. The results of this paper is expected to provide input in the preparation of GR on licensing of nuclear installations. (author)

Review of the Brazilian experience in the licensing of nuclear power plants

International Nuclear Information System (INIS)

Lederman, L.; Laborne, J.J.

1983-01-01

Survey of the licensing of the Brazilian Nuclear Power Plants (NPPs) is presented. The organization and technical expertise of the Comissao Nacional de Energia Nuclear, the Brazilian Regulatory Body, is reviewed with regard to in-house experience, foreign consultants, agreements with regulatory bodies of other countries and research contracts with Brazilian univerisities. The application of the two-stage licensing process and the stage of development of Brazilian nuclear standards is described. Finally, the paper speculates about the future role of probabilistic risk assessment in the Brazilian licensing process

RHTF 2, a 1200 MWe high temperature reactor

International Nuclear Information System (INIS)

Brisbois, Jacques

1978-01-01

After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

Gas-cooled reactors

International Nuclear Information System (INIS)

Schulten, R.; Trauger, D.B.

1976-01-01

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

Finnish experiences on licensing and using of programmable digital systems in nuclear power plants

International Nuclear Information System (INIS)

Haapanen, P.; Maskuniitty, M.; Heimburger, H.; Hall, L.E.; Manninen, T.

1993-01-01

Finnish utility companies, Imatran Voima Oy (IVO) and Teollisuuden Voima (TVO), and the licensing authority, the Finnish Centre for Radiation and Nuclear Safety (STUK), are preparing for a new nuclear power plant in Finland. Plant vendors are proposing programmable digital automation systems for both the safety-related and the operational I and C (instrumentation and control) systems in this new unit. Also in existing plant units the replacement of certain old analog systems with state-of-the-art digital ones will become necessary in the years to come. Licensing of programmable systems for safety critical applications requires a new approach due to the special properties and failure modes of these systems. The major difficulties seem to be in the assessment and quantification of software reliability. The Technical Research Centre of Finland has in co-operation with the authority and the utilities conducted a project (AJA) to develop domestically applicable licensing requirements, guidelines and practices. International standards, guidelines and licensing practices have been analyzed in order to specify national licensing requirements. The paper describes and discusses the findings and experiences of the AJA project so far. The experience in introducing advanced programmable digital control and computer systems in the operating nuclear power plants will be covered briefly. Although these systems are not safety-related but systems of more general interest regarding nuclear safety, some routines regarding the licensing of safety- related systems have been followed. In these backfitting and replacement projects some experience have been gained in how to license safety-related programmable systems. (Author) 31 refs., 2 figs

Results of the Nucifer reactor neutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

2016-07-01

Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

Power reactor events, May-June 1986

International Nuclear Information System (INIS)

Massaro, S.A.

1986-12-01

Power Reactor Events is a bi-monthly newsletter that compiles operating experience information about commercial nuclear power plants. This includes summaries of noteworthy events and listings and/or abstracts of USNRC and other documents that discuss safety-related or possible generic issues. It is intended to feed back some of the lessons learned from operational experience to the various plant personnel, i.e., managers, licensed reactor operators, training coordinators, and support personnel. Events at the following plants are reported: McGuire Unit 1; Susquehanna Units 1 and 2; Browns Ferry Units 1, 2, and 3; and River Bend Unit 1

Selection/licensing of nuclear power plant operators

International Nuclear Information System (INIS)

Saari, L.M.

1983-07-01

An important aspect of nuclear power plant (NPP) safety is the reactor operator in the control room. The operators are the first individuals to deal with an emergency situation, and thus, effective performance on their part is essential for safe plant operations. Important issues pertaining to NPP reactor operators would fall within the personnel subsystem of our safety system analysis. While there are many potential aspects of the personnel subsystem, a key first step in this focus is the selection of individuals - attempting to choose individuals for the job of reactor operator who will safely perform the job. This requires a valid (job-related) selection process. Some background information on the Nuclear Regulatory Commission (NRC) licensing process used for selecting NPP reactor operators is briefly presented and a description of a research endeavor now underway at Battelle for developing a valid reactor operator licensing examination is included

Licensed operating reactors: Status summary report, data as of December 31, 1995. Volume 20

International Nuclear Information System (INIS)

1996-06-01

The US Nuclear Regulatory Commission's monthly summary of licensed nuclear power reactor data is based primarily on the operating data report submitted by licensees for each unit. This report is divided into two sections: the first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 availability factors, capacity factors, and forced outage rates are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensees and notes to the use of weighted averages and starting dates other than commercial operation are provided

Operating experience with the DRAGON High Temperature Reactor experiment

International Nuclear Information System (INIS)

Simon, R.A.; Capp, P.D.

2002-01-01

The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

CONTAIN 2.0 code release and the transition to licensing

International Nuclear Information System (INIS)

Murata, K.K.; Griffith, R.O.; Bergeron, K.D.; Tills, J.

1997-10-01

CONTAIN is a reactor accident simulation code developed by Sandia National Laboratories under US Nuclear Regulatory Commission (USNRC) sponsorship to provide integrated analysis of containment phenomena, including those related to nuclear reactor containment loads and radiological source terms. The recently released CONTAIN 2.0 code version represents a significant advance in CONTAIN modeling capabilities over the last major code release (CONTAIN 1.12V). The new modeling capabilities are discussed here. The principal motivation for many of the recent model improvements has been to allow CONTAIN to model the special features in advanced light water reactor (ALWR) designs. The work done in this area is also summarized. In addition to the ALWR work, the USNRC is currently engaged in an effort to qualify CONTAIN for more general use in licensing, with the intent of supplementing or possibly replacing traditional licensing codes. To qualify the CONTAIN code for licensing applications, studies utilizing CONTAIN 2.0 are in progress. A number of results from this effort are presented in this paper to illustrate the code capabilities. In particular, CONTAIN calculations of the NUPEC M-8-1 and ISP-23 experiments and CVTR test number-sign 3 are presented to illustrate (1) the ability of CONTAIN to model non-uniform gas density and/or temperature distributions, and (2) the relationship between such gas distributions and containment loads. CONTAIN and CONTEMPT predictions for a large break loss of coolant accident scenario in the San Onofre plant are also compared

Reactor licensing process: a status report

International Nuclear Information System (INIS)

Long, J.A.

1977-01-01

The Nuclear Regulatory Commission (NRC), in its review of applications for licenses to construct and operate nuclear power plants, is required to consider those measures necessary to ensure the protection of the health and safety of the public and the environment. The article discusses the NRC staff procedures and policies for conducting the detailed safety, environmental, and antitrust reviews that provide the basis for these assurances. Included is a discussion of the improvements to the licensing process currently being proposed or implemented to enhance its stability and predictability for the benefit of all involved with the regulation of nuclear power. The views and opinions expressed in the article are those of the author alone and do not represent positions of the NRC

International working group on gas-cooled reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

1981-01-15

The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

Safety Evaluation Report related to the renewal of the operating license for the General Electric-Nuclear Test Reactor (GE-NTR) (Docket No. 50-73)

International Nuclear Information System (INIS)

1984-09-01

This Safety Evaluation Report for the application filed by the General Electric Company (GE) for a renewal license number R-33 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by GE and is located in Pleasanton, California. The staff concludes that the reactor can continue to be operated by GE without endangering the health and safety of the public

Safety Evaluation Report related to the renewal of the operating license for the research reactor at Manhattan College (Docket No. 50-199)

International Nuclear Information System (INIS)

1985-02-01

This Safety Evaluation Report for the application filed by Manhattan College (MC) for a renewal of Operating License R-94 to continue to operate the MC 0.1-W open-pool training reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by MC and is located two blocks away from the MC main campus in the Riverdale area of New York City, New York. The staff concludes that the reactor facility can continue to be operated by MC without endangering the health and safety of the public

Safety evaluation report related to the renewal of the operating license for the University of New Mexico Research Reactor (Docket No. 50-252)

International Nuclear Information System (INIS)

1987-03-01

This Safety Evaluation Report for the application filed by the University of New Mexico (UNM) for renewal of Operating License No. R-102 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the campus of the University of New Mexico in Albuquerque, New Mexico. The staff concludes that the reactor can continue to be operated by the University of New Mexico without endangering the health and safety of the public. 7 refs., 7 figs., 2 tabs

Licensed operating reactors. Status summary report data as of March 31, 1986. Volume 10, No. 4

International Nuclear Information System (INIS)

1986-05-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States. The percentage computations, Items 20 through 24 in Section 2, the vendor capacity factors on page 1-7, and actual vs potential energy production on page 1-2 are computed using actual data for the period of consideration. The percentages listed in power generation on page 1-2 are computed as a arithmetic average. The factors for the life-span of each unit (the ''Cumulative'' column) are reported by the utility and are not entirely recomputed by NRC. Utility power production data is checked for consistency with previously submitted statistics. It is hoped this status report proves informative and helpful to all agencies and individuals interested in analyzing trends in the nuclear industry which might have safety implications, or in maintaining an awareness of the US energy situation as a whole

IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

International Nuclear Information System (INIS)

Sartori, Enrico; Blair Briggs, J.

2008-01-01

1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

A licensing discussion: SMRs in Canada

International Nuclear Information System (INIS)

De Vos, M.

2013-01-01

The CNSC (Canadian AECB) is ready to regulate Small Modular Reactors (SMR) facilities in Canada. The CNSC is well situated to engage with proponents of SMR reactors in design reviews or licensing discussions. A risk-informed (graded) approach is possible in many instances for reactors but it is not a relaxation of requirements. The vendor design review process helps reduce regulatory risks by encouraging early engagement.

Licensed operating reactors. Status summary report data as of December 31, 1993

Energy Technology Data Exchange (ETDEWEB)

Hartfield, R.A.

1994-03-01

The Nuclear Regulatory Commissions annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December, the year to date (in this case calendar year 1993) and cumulative data, usually for the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided.

Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

Energy Technology Data Exchange (ETDEWEB)

Moe, Wayne L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-08-23

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

Energy Technology Data Exchange (ETDEWEB)

Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

Lessons from early experience in reactor development

International Nuclear Information System (INIS)

Allen, W.

1976-09-01

This paper deals with several issues in U.S. reactor development and demonstration experience. The focus is on the period between 1946 and 1963 during which the Atomic Energy Commission (AEC) guided early reactor research and development (R and D) and conducted the Power Reactor Demonstration Program

International exchange of safety and licensing information

International Nuclear Information System (INIS)

Lafleur, J.D. Jr.; Hauber, R.D.; Chenier, D.M.

1977-01-01

A network of formal and informal bilateral arrangements for the exchange of nuclear safety information is being established by the U.S. Nuclear Regulatory Commission. For developing countries, such arrangements can provide ready access to the extensive, fully documented safety analyses and safety research results that NRC has accumulated. NRC has been receiving foreign visitors at a rate of about 500 per year, largely for discussions of safety and licensing questions related to light water reactors. Exchanges also are taking place on the safety of advanced reactors. A special interest of the NRC is in providing for reciprocal communicaion, at the earliest possible time, of important problems, decisions and other actions on nuclear safety matters. For example, it is essential that a newly-discovered problem in a nuclear reactor be brought immediately to the attention of other governments which are responsible for the safety of similar reactors. Definite progress has been made in the U.S. Freedom of Information Act. Certain exchanges have taken place on this basis. Experience in the establishment and operation of NRC's bilateral exchange arrangements is summarized. A typical exchange with the regulatory authority of country building its first power reactor is described

Management of spent fuel from power and research reactors using CASTOR and CONSTOR casks and licensing experience worldwide

International Nuclear Information System (INIS)

Becher, D.

2003-01-01

An overview of the spent fuel storage in CASTOR and CONSTOR casks during the last 30 years is made. Design characteristics of the both types of casks are presented. CASTOR casks fulfill both the requirements for type B packages according to the IAEA requirements covering different accident situations in storage sites. Analyses of nuclear and thermal behavior and strength are carried out for CONSTOR concept. Special experimental program for verification of mechanical and thermomechanical properties is implemented. Licensing experience of the casks in German storage facilities is presented. Special modifications of CASTOR casks for WWER-440 and RBMK fuel assemblies have been designed for implementation in Eastern Europe. Contracts for GNB spent fuel casks delivery are concluded with Czech Republic, Slovakia, Hungary and Lithuania

Final Environmental Statement related to license renewal and power increase for the National Bureau of Standards Reactor: Docket No. 50-184

International Nuclear Information System (INIS)

1982-08-01

This Final Environmental Statement contains an assessment of the environmental impact associated with renewal of Operating License No. TR-5 for the National Bureau of Standards (NBS) reactor for a period of 20 years at a power level of 20 MW. This reactor is located on the 576-acre NBS site near Gaithersburg in Montgomery County, Maryland, about 20 mi northwest of the center of Washington, DC. The reactor is a high-flux heavy-water-moderated, cooled and reflected test reactor, which first went critical on December 7, 1967. Though the reactor was originally designed for 20-MW operation, it has been operating for 14 years at a maximum authorized power level to 10 MW. Program demand is now great enough to warrant operation at a power level of 20 MW. No additional major changes to the physical plant are required to operate at 20 MW

Concept study for interim storage of research reactor fuel elements in transport and storage casks. Transport and storage licensing procedure for the CASTOR MTR 2 cask. Final report

International Nuclear Information System (INIS)

Weiss, M.

2001-01-01

As a result of the project, a concept was to be developed for managing spent fuel elements from research reactors on the basis of the interim storage technology existing in Germany, in order to make the transition to direct disposal possible in the long term. This final report describes the studies for the spent fuel management concept as well as the development of a transport and storage cask for spent fuel elements from research reactors. The concept analyses were based on data of the fuel to be disposed of, as well as the handling conditions for casks at the German research reactors. Due to the quite different conditions for handling of casks at the individual reactors, it was necessary to examine different cask concepts as well as special solutions for loading the casks outside of the spent fuel pools. As a result of these analyses, a concept was elaborated on the basis of a newly developed transport and storage cask as well as a mobile fuel transfer system for the reactor stations, at which a direct loading of the cask is not possible, as the optimal variant. The cask necessary for this concept with the designation CASTOR trademark MTR 2 follows in ist design the tried and tested principles of the CASTOR trademark casks for transport and interim storage of spent LWR fuel. With the CASTOR trademark MTR 2, it is possible to transport and to place into long term interim storage various fuel element types, which have been and are currently used in German research reactors. The technical development of the cask has been completed, the documents for the transport license as type B(U)F package design and for obtaining the storage license at the interim storage facility of Ahaus have been prepared, submitted to the licensing authorities and to a large degree already evaluated positively. The transport license of the CASTOR trademark MTR 2 has been issued for the shipment of VKTA-contents and FRM II compact fuel elements. (orig.)

New nuclear plant design and licensing process

International Nuclear Information System (INIS)

Luangdilok, W.

1996-01-01

This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

The experiences of research reactor accident to safety improvement

International Nuclear Information System (INIS)

Wiranto, S.

1999-01-01

The safety of reactor operation is the main factor in order that the nuclear technology development program can be held according the expected target. Several experience with research reactor incidents must be learned and understood by the nuclear program personnel, especially for operators and supervisors of RSG-GA. Siwabessy. From the incident experience of research reactor in the world, which mentioned in the book 'Experience with research reactor incidents' by IAEA, 1995, was concluded that the main cause of research reactor accidents is understandless about the safety culture by the nuclear installation personnel. With learn, understand and compare between this experiences and the condition of RSG GA Siwabessy is expended the operators and supervisors more attention about the safety culture, so that RSG GA Siwabessy can be operated successfull, safely according the expected target

Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

International Nuclear Information System (INIS)

1978-09-01

The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

Energy Technology Data Exchange (ETDEWEB)

Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

1999-08-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

International Nuclear Information System (INIS)

Hastowo, Hudi; Tarigan, Alim

1999-01-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Safety-evaluation report related to the license renewal and power increase for the National Bureau of Standards Reactor (Docket No. 50-184)

International Nuclear Information System (INIS)

1983-09-01

This Safety Evaluation Report for the application filed by the National Bureau of Standards (NBS) for an increase in power from 10 MWt to 20 MWt and for a renewal of the Operating License TR-5 to continue to operate the test reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Gaithersburg, Maryland, on the site of the National Bureau of Standards, which is a bureau of the Department of Commerce. The staff concludes that the NBS reactor can operate at the 20 MWt power level without endangering the health and safety of the public

Safety evaluation report related to the license renewal and power increase for the National Bureau of Standards reactor (Docket No. 50-184)

International Nuclear Information System (INIS)

Bernard, H.

1984-03-01

Supplement 1 to the Safety Evaluation Report (SER) related to the renewal of the operating license and for a power increase (10 MWt to 20 MWt) for the research reactor at the National Bureau of Standards (NBS) facility has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the review of the licensee's emergency plan, which had not been reviewed at the time the Safety Evaluation Report (NUREG-1007) was published, and the review of the NBS application by the Advisory Committee on Reactor Safeguards, which was completed subsequent to the publication of the SER

Upgrading of the research reactors FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1981-01-01

In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

Safety evaluation report related to the renewal of the operating license for the research reactor at the University of Florida. Docket No. 50-83

International Nuclear Information System (INIS)

1982-05-01

This Safety Evaluation Report for the application filed by the University of Florida (UF) for a renewal of Operating License R-56 to continue to operate its Argonaut-type research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the University of Florida and is located on the UF campus in Gainesville, Alachua County, Florida. The staff concludes that the reactor facility can continue to be operated by UF without endangering the health and safety of the public

Safety Evaluation Report related to the renewal of the operating license for the research reactor at Michigan State University (Docket No. 50-294)

International Nuclear Information System (INIS)

1984-08-01

This Safety Evaluation Report for the application filed by the Michigan State University (MSU) for a renewal of operating license number R-114 to continue to operate the TRIGA Mark I research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Michigan State University and is located on the campus of Michigan State University in East Lansing, Ingham County, Michigan. The staff concludes that the TRIGA reactor facility can continue to be operated by MSU without endangering the health and safety of the public

PBMR-SA licensing project organization

International Nuclear Information System (INIS)

Clapisson, G.A.; Metcalf, P.E.; Mysen, A.

2001-01-01

The South African nuclear regulatory authority, the Council for Nuclear Safety (CNS), is beginning the safety review of the Pebble Bed Modular Reactor (PBMR) design under development by the South African National Electrical Utility, Eskom. This paper describes the CNS licensing process, including the establishment of basic licensing criteria, general design criteria, and specific design rules, as well the safety assessment to be conducted in accordance with the established structure. It also summarises the CNS PBMR review project activities, including the overall organisational arrangements, licensing basis, safety and risk assessment, general operating rules and plant design engineering, and pre-operational testing. (author)

Safety Evaluation Report related to the renewal of the operating license for the training and research reactor at the University of Michigan (Docket No. 50-2)

International Nuclear Information System (INIS)

1985-07-01

This Safety Evaluation Report for the application filed by the University of Michigan (UM) for renewal of the Ford Nuclear Reactor (FNR) operating license number R-28 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the North Campus of the University of Michigan in Ann Arbor, Michigan. The staff concludes that the reactor can continue to be operated by the University of Michigan without endangering the health and safety of the public

Operational and reliability experience with reactor instrumentation

International Nuclear Information System (INIS)

Dixon, F.; Gow, R.S.

1978-01-01

In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

Study on Operator Actions during the Occurrences of Undesirable Events in PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Tom, P.P.; Nurul Husna Zainal Abidin; Lanyau, T.A.; Zaredah Hashim

2016-01-01

Due to the recent Fukushima accident, the potential risks at one and only nuclear research reactor in the country, which is the PUSPATI TRIGA Reactor (RTP), has increasingly gain concerns and an attempt on the development of Level 1 Probabilistic Safety Assessment (PSA) for this reactor has been commenced. The preliminary scope of the PSA is to analyse the risk of core degradation during normal daily operation due to the random component failure and human error. SPAR-H and THERP method is used for quantifying human error probability (HEP). However, the scopes of this study only cover the qualitative parts that use interview/questionnaire method. The objectives of the questionnaire are to identify the main action for RTP operators when any undesired incident occurs during full power operation that might be caused by random component failures. From the questionnaires that have been conducted, the respondents consisted of 4 licensed operators and 9 trainee operators. All licensed operators have experience of operating reactor for more than 15 years while the trainee operator have been operate the reactor with experience of less than 10 years. Generally, in the event of an abnormal condition involving the reactor, an operator whether a licensed operator or the trainee does not have to ask permission in advance from the top individuals to carry out scram. This is to prevent the situation becoming increasingly severe if the reactor is still operating. With complete training and knowledge derived from the management, an operator can act efficiently in any emergency case. (author)

Management of the licensing of users of radioactive materials should be improved

International Nuclear Information System (INIS)

1976-01-01

Radioactive material licenses are required for manufacturing nuclear fuel for reactors and for industrial, commercial, medical, and educational uses of radioactive materials. This type of license is not for constructing or operating nuclear power reactors and facilities for processing used nuclear fuels. This report discusses the need for better management improvements in the NRC's program for licensing the users. As of December 31, 1974, there were 8,253 active NRC-issued material licenses held by 6,310 licensees. The study reviewed NRC's policies, procedures, and practices, and examined recent evaluations of state programs to identify problems encountered by the states

Atomic Safety and Licensing Board Panel annual report

International Nuclear Information System (INIS)

1991-09-01

In Fiscal Year 1990, The Atomic Safety and Licensing Board Panel (Panel) handled 40 proceedings involving the construction, operation, and maintenance of commercial nuclear power reactors or other activities requiring a license from the Nuclear Regulatory Commission. This report summarizes, highlights, and analyzes how the judges and licensing boards of the Panel addressed the wide-ranging issues raised in these proceedings during the year

Licensed operating reactors: Status summary report data as of December 31, 1992

International Nuclear Information System (INIS)

Hartfield, R.A.

1993-03-01

The Nuclear Regulatory Commission's annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar year 1992) and cumulative data, usually from the date of commercial operation. ne data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

NRC Licensing Status Summary Report for NGNP

Energy Technology Data Exchange (ETDEWEB)

Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, James Carl [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-11-01

The Next Generation Nuclear Plant (NGNP) Project, initiated at Idaho National Laboratory (INL) by the U.S. Department of Energy (DOE) pursuant to provisions of the Energy Policy Act of 2005, is based on research and development activities supported by the Department of Energy Generation IV Nuclear Energy Systems Initiative. The principal objective of the NGNP Project is to support commercialization of high temperature gas-cooled reactor (HTGR) technology. The HTGR is a helium-cooled and graphite moderated reactor that can operate at temperatures much higher than those of conventional light water reactor (LWR) technologies. The NGNP will be licensed for construction and operation by the Nuclear Regulatory Commission (NRC). However, not all elements of current regulations (and their related implementation guidance) can be applied to HTGR technology at this time. Certain policies established during past LWR licensing actions must be realigned to properly accommodate advanced HTGR technology. A strategy for licensing HTGR technology was developed and executed through the cooperative effort of DOE and the NRC through the NGNP Project. The purpose of this report is to provide a snapshot of the current status of the still evolving pre-license application regulatory framework relative to commercial HTGR technology deployment in the U.S. The following discussion focuses on (1) describing what has been accomplished by the NGNP Project up to the time of this report, and (2) providing observations and recommendations concerning actions that remain to be accomplished to enable the safe and timely licensing of a commercial HTGR facility in the U.S.

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

International Nuclear Information System (INIS)

Moe, Wayne Leland

2015-01-01

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a ''critical path'' for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain ''minimum'' levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial ''first step'' in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by

Research and development associated with licensing of MHTGR [Modular High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Jones, H.

1990-01-01

The Modular High Temperature Gas-Cooled Reactor (MHTGR) currently under development by the US Department of Energy (US-DOE) for commercial applications has top-level goals of producing safe, economical power for the US utility industry. The utility industry has been represented in formulating design and licensing requirements through both a ''Utility User Requirements Document'' and by participating in the DOE system engineering process known as the ''Integrated Approach.'' The result of this collaboration has been to set stringent goals for both the safety and operational reliability of the MHTGR. To achieve these goals, the designer must have access to a more comprehensive data base of properties in several fields of technology than is currently available. A technology development program has been planned to provide this data to the designer in time to support both his design activities and the submittal of formal licensing application documents. The US-DOE has chosen the Oak Ridge National Laboratory (ORNL) to take the lead in planning and executing these technology programs. When completed these will augment the designer's current data base and provide the necessary depth to meet the stringent goals which have been set for the MHTGR. It is worth noting that the goals of safety and operational reliability are complementary, and the data required from the technology development program will be similar. Therefore, the program to support the licensing of the MHTGR is not separate from that required for design, but is a subset of that which meets all the requirements that result from implementing the US-DOE's integrated approach. 38 figs

75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Science.gov (United States)

2010-11-16

... NUCLEAR REGULATORY COMMISSION [Docket Nos (Redacted), License Nos (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I. The licensees identified in...

75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Science.gov (United States)

2010-12-20

... NUCLEAR REGULATORY COMMISSION [Docket Nos. (Redacted), License Nos.: (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I The licensees identified in...

Safety Evaluation Report, related to the renewal of the operating license for the critical experiment facility of the Rensselaer Polytechnic Institute (Docket No. 50-225)

International Nuclear Information System (INIS)

1983-10-01

This Safety Evaluation Report for the application filed by the Rensselaer Polytechnic Institute (RPI) for a renewal of operating license CX-22 to continue to operate a critical experiment facility has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by Rensselaer Polytechnic Institute and is located at a site in the city of Schenectady, New York. The staff concludes that this critical facility can continue to be operated by RPI without endangering the health and safety of the public

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 ⁷LiF-BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

International Nuclear Information System (INIS)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

2016-01-01

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

The reactor antineutrino anomaly and low energy threshold neutrino experiments

Science.gov (United States)

Cañas, B. C.; Garcés, E. A.; Miranda, O. G.; Parada, A.

2018-01-01

Short distance reactor antineutrino experiments measure an antineutrino spectrum a few percent lower than expected from theoretical predictions. In this work we study the potential of low energy threshold reactor experiments in the context of a light sterile neutrino signal. We discuss the perspectives of the recently detected coherent elastic neutrino-nucleus scattering in future reactor antineutrino experiments. We find that the expectations to improve the current constraints on the mixing with sterile neutrinos are promising. We also analyze the measurements of antineutrino scattering off electrons from short distance reactor experiments. In this case, the statistics is not competitive with inverse beta decay experiments, although future experiments might play a role when compare it with the Gallium anomaly.

Current safety issues of CANDU licensing

International Nuclear Information System (INIS)

Lee, Y.; Natalizio, A.

1994-01-01

As requested by Korea Institute of Nuclear Safety(KINS), the status of five generic licensing issues has been examined and their potential impact on a new plant that would be constructed in Canada has been evaluated. The results and conclusions of this evaluation are summarized as follows: steam explosion in calandria, hydrogen explosion in containment, use of PSA in reactor licensing, human factors, safety critical software

Storage of water reactor spent fuel in water pools. Survey of world experience

International Nuclear Information System (INIS)

1982-01-01

Following discharge from a nuclear reactor, spent fuel has to be stored in water pools at the reactor site to allow for radioactive decay and cooling. After this initial storage period, the future treatment of spent fuel depends on the fuel cycle concept chosen. Spent fuel can either be treated by chemical processing or conditioning for final disposal at the relevant fuel cycle facilities, or be held in interim storage - at the reactor site or at a central storage facility. Recent forecasts predict that, by the year 2000, more than 150,000 tonnes of heavy metal from spent LWR fuel will have been accumulated. Because of postponed commitments regarding spent fuel treatment, a significant amount of spent fuel will still be held in storage at that time. Although very positive experience with wet storage has been gained over the past 40 years, making wet storage a proven technology, it appears desirable to summarize all available data for the benefit of designers, storage pool operators, licensing agenices and the general public. Such data will be essential for assessing the viability of extended water pool storage of spent nuclear fuel. In 1979, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD jointly issued a questionnaire dealing with all aspects of water pool storage. This report summarizes the information received from storage pool operators

Licensed operating reactors. Status summary report, data as of April 30, 1986

International Nuclear Information System (INIS)

1986-05-01

This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report data as of 06-30-86

International Nuclear Information System (INIS)

1986-11-01

This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Safety Evaluation Report related to renewal of the operating license for the CAVALIER Training Reactor at the University of Virginia (Docket No. 50-396)

International Nuclear Information System (INIS)

1985-05-01

This Safety Evaluation Report for the application filed by the University of Virginia for a renewal of Operating License R-123 to continue to operate the CAVALIER (Cooperatively Assembled Virginia Low Intensity Educational Reactor) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the University of Virginia and is located on the campus in Charlottesville, Virginia. Based on its technical review, the staff concludes that the reactor facility can continue to be operated by the university without endangering the health and safety of the public or the environment

Safety evaluation report related to the renewal of the operating license for the Worcester Polytechnic Institute open-pool training reactor, Docket No. 50-134

International Nuclear Information System (INIS)

1982-12-01

This Safety Evaluation Report for the application filed by the Worcester Polytechnic Institute (WPI) for a renewal of Operating License R-61 to continue to operate the WPI 10-kW open-pool training reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Worcester Polytechnic Institute and is located on the WPI campus in Worcester, Worcester County, Massachusetts. The staff concludes that the reactor facility can continue to be operated by WPI without endangering the health and safety of the public

Safety Evaluation Report related to the renewal of the operating license for the research reactor at the University of Kansas (Docket No. 50-148)

International Nuclear Information System (INIS)

1984-05-01

This Safety Evaluation Report for the application filed by the University of Kansas (KU) for a renewal of Operating License R-78 to continue to operate the KU 250-kW open-pool training reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the University of Kansas and is located on the KU campus in Lawrence, Douglas County, Kansas. The staff concludes that the reactor facility can continue to be operated by KU without endangering the health and safety of the public. 17 references, 11 figures, 4 tables

Discussion on safety analysis approach for sodium fast reactors

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Suh, Nam Duk; Shin, Ahn Dong; Bae, Moo Hoon

2012-01-01

above paragraphs the conventional deterministic approach has some defects in directly applying to SFR mainly because of its reactor specific framework and the complete risk informed approach (TNF) has not been verified sufficiently with much uncertainty. This study presents some insight from critical review of the deterministic approach, risk informed approach, recent trend in each country, and past SFR licensing approach experiences. The insights are expected to give useful guide for the setup of licensing approach in blending safety analysis approach in stand of deterministic position for the urgent application of the approach

RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1984-12-01

During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization

RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1985-01-01

According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

Safety Evaluation Report related to the renewal of the operating license for the training and research reactor at the University of Lowell (Docket No. 50-223)

International Nuclear Information System (INIS)

1985-11-01

This Safety Evaluation Report for the application filed by the University of Lowell (UL) for renewal of operating license number R-125 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the North Campus of the University of Lowell in Lowell, Massachusetts. The staff concludes that the reactor can continue to be operated by the University of Lowell without endangering the health and safety of the public

International Reactor Physics Experiment Evaluation (IRPhE) Project

International Nuclear Information System (INIS)

2013-01-01

The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

CANDU reactor experience: fuel performance

International Nuclear Information System (INIS)

Truant, P.T.; Hastings, I.J.

1985-07-01

Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

Different types of power reactors and provenness

International Nuclear Information System (INIS)

Goodman, E.I.

1977-01-01

The lecture guides the potential buyer in the selection of a reactor type. Recommended criteria regarding provenness, licensability, and contractual arrangements are defined and discussed. Tabular data summarizing operating experience and commercial availability of units are presented and discussed. The status of small and medium power reactors which are of interest to many developing countries is presented. It is stressed that each prospective buyer will have to establish his own criteria based on specific conditions which will be applied to reactor selection. In all cases it will be found that selection, either pre-selection of bidders or final selection of supplier, will be a fairly complex evaluation. (orig.) [de

Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

International Nuclear Information System (INIS)

Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

1996-09-01

The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

Containment atmosphere cooling system for experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

1979-01-01

The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

Experiences in stability testing of boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Otaduy, P.J.

1986-01-01

The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

Licensing of nuclear and radioactive installations in Peru

International Nuclear Information System (INIS)

Medina Gironzini, E.

1987-01-01

In Peru, the Regulation for Ionizing Radiation Sources is applied, which establishes the norms and procedures to follow in the nuclear and radioactive installations of the country in order to assure their correct operation as concerns to the nuclear safety and radiological protection, allowing the emission of the respective licenses. As for the nuclear facilities, this authorization includes the Previous License, the Construction License and the Operation License (provisional and definitive) and for radioactive facilities and equipment generating ionizing radiations: the Construction License and the Operation License. The personnel also require a license that can be an operator license (as for nuclear reactors) or a supervisor license (for nuclear and radioactive facilities). In spite of the above mentioned regulation and its long enforcement period, less than 10% of radioactive facilities in this country are licensed, due to different problems which will be solved in the medium term. (Author)

Licensing process for the power increase

International Nuclear Information System (INIS)

Perez P, R.; Mamani A, Y.R.; Salgado G, J.R.

2006-01-01

The licensing process for the power increase of a nuclear power station is presented, this includes the description of the effective normative framework, the attributions of the one Mexican regulator organism in nuclear matter, the definition and importance of the power level of a nuclear reactor for the safety studies. Also, the types of power increase according to its magnitude, and the regulator process that it includes the scope and the detail of the required information that it should be evaluated by the one regulator organism are discussed. Finally it offers a summary of the experience that one has in Mexico for this type of processes. (Author)

The road to MYRRHA. The pre-licensing phase and FEED

International Nuclear Information System (INIS)

2014-01-01

The article discusses the study phase of the future research reactor MYRRHA. This study phase covers three major areas: Research and Development, licensing and engineering. In 2013, substantial progress was made in the licensing and engineering processes: licensing is currently in the pre-licensing phase, while, for engineering, the Front-end Engineering and Design (FEED) contract was signed for the construction of the non-nuclear parts of MYRRHA. The article highlights work associated with the the pre-licensing phase and the FEED.

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Experiment operations plan for the MT-4 experiment in the NRU reactor

International Nuclear Information System (INIS)

Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

1983-06-01

A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700 0 F)

Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Reactor safety assessment system

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Reactor Safety Research: Semiannual report, July-December 1986

Energy Technology Data Exchange (ETDEWEB)

1987-11-01

Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

Reactor Safety Research: Semiannual report, July-December 1986

International Nuclear Information System (INIS)

1987-11-01

Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions

NRC licensing requirements: DOD options

International Nuclear Information System (INIS)

Pike, W.J.; O'Reilly, P.D.

1982-09-01

This report describes the licensing process (both safety and environmental) that would apply if the Department of Defense (DOD) chooses to obtain licenses from the US Nuclear Regulatory Commission (NRC) for using nuclear energy for power and luminous sources. The specific nuclear energy sources being considered include: small or medium-size nuclear power reactors; radioisotopic thermoelectric generators with 90 Sr or 238 Pu; radioisotopic dynamic electric generators with 90 Sr or 238 Pu; and applications of radioisotopes for luminous sources (lights) with 3 H, 85 Kr, or 147 Pm. The steps of the licensing process are summarized in the following sections, with particular attention given to the schedule and level of effort necessary to support the process

Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Soyama, Kazuhiko; Ichikawa, Hiroki

1991-08-01

According to a reduction of fuel enrichment from 45 w/o 235 U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm 3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO 2 -zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

Safety Evaluation Report related to the renewal of the operating license for the research reactor at the Universiy of Missouri-Rolla (Docket No. 5-123)

International Nuclear Information System (INIS)

1984-12-01

This Safety Evaluation Report for the application filed by the University of Missouri-Rolla for a renewal of Operating License R-79 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned by the University of Missouri and is located on the campus in Rolla, Missouri. On the basis of its technical review, the staff concludes that the reactor facility can continue to be operated by the university without endangering the health and safety of the public or the environment

Experience in arranging shipments of spent fuel assemblies of commercial and research reactors

International Nuclear Information System (INIS)

Komarov, S.; Barinkov, O.; Eshcherkin, A.; Lozhnikov, V.; Smirnov, A.

2008-01-01

At present the key activities of Sosny Company are to inspect physical conditions, handle and arrange shipment of SFA including failed SFA. In 2003 after obtaining the license of Gosatomnadzor (Rostechnadzor now) entitled to handle nuclear materials in the process of their shipment, Sosny Company started preparing certification and arranging SFA shipment on its own. About 40 shipments of SFA were performed with participation of Sosny Company. Experience in handling failed SFA - an example of development of a new technology could be the transport and technological scheme of RBMK-1000 SFA shipment from Leningradskaya NPP that was designed by Sosny Company. TUK-11 cask was selected for this shipment. The example of change of transport and technological scheme is modification of the technology for handling and shipment of WWER-440 SFA from Kola NPP. Experience in arranging transportation - based on the results of development of logistics schemes for shipping SFA of reactor facilities Sosny Company justified and implemented composition of mixed trains containing rail cars of many types that enabled to perform shipment more efficiently in time and cost. Experience in arranging handling and shipment of research reactor SFA - over the past years the activity of Sosny Company was aimed at implementing international Russian Research Reactor Fuel Return (RRRFR) program. Since equipment of the majority of research centers doesn't allow for the large casks to be accepted and loaded, special casks of less mass and dimensions are used to ship SFA from research reactors. In RRRFR program it is assumed to use different casks for RR SFA such as Russian TUK- 19, TUK-128 and foreign SKODA VPVR/M and NAC-LWT. At present Sosny Company is involved in coordination of the efforts of the affected organizations in creating the type 'C' package for RR SFA in the RF. Conclusion: Under conditions of constant increase of the requirements to shipment safety and complication of regulations of all

Safety Evaluation Report related to the construction permit and operating license for the research reactor at the University of Texas (Docket No. 50-602)

International Nuclear Information System (INIS)

1985-05-01

This Safety Evaluation Report for the application filed by the University of Texas for a construction permit and operating license to construct and operate a TRIGA research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the University of Texas and is located at the university's Balcones Research Center, about 7 miles (11.6 km) north of the main campus in Austin, Texas. The staff concludes that the TRIGA reactor facility can be constructed and operated by the University of Texas without endangering the health and safety of the public

From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

Safety evaluation report related to the renewal of the operating license for the Zero-Power Reactor at Cornell University, Docket No. 50-97

International Nuclear Information System (INIS)

1983-09-01

This Safety Evaluation Report for the application filed by Cornell University (CU) for a renewal of Operating License R-80 to continue to operate a zero-power reactor (ZPR) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by Cornell University and is located on the Cornell campus in Ithaca, New York. The staff concludes that the ZPR facility can continue to be operated by CU without endangering the health and safety of the public

ITER licensing

International Nuclear Information System (INIS)

Gordon, C.W.

2005-01-01

ITER was fortunate to have four countries interested in ITER siting to the point where licensing discussions were initiated. This experience uncovered the challenges of licensing a first of a kind, fusion machine under different licensing regimes and helped prepare the way for the site specific licensing process. These initial steps in licensing ITER have allowed for refining the safety case and provide confidence that the design and safety approach will be licensable. With site-specific licensing underway, the necessary regulatory submissions have been defined and are well on the way to being completed. Of course, there is still work to be done and details to be sorted out. However, the informal international discussions to bring both the proponent and regulatory authority up to a common level of understanding have laid the foundation for a licensing process that should proceed smoothly. This paper provides observations from the perspective of the International Team. (author)

The current role of simulators for performance evaluations and licensing (case of Mexico)

International Nuclear Information System (INIS)

Maldonado A, H.

1997-01-01

The main purpose of this paper is to share the experience acquired by the National Commission of Nuclear Safety and Safeguards (CNSNS) during the administration of both certification and licensing operational exams applied to Senior Reactor Operators (SRO) and Reactor OPerators (RO) by using a full-scope simulator. The licensing operational exams are administered to examinate candidates for a SRO or RO license while the certification operational exams are administered to all personnel that possess a SRO or RO license in order to renew their licenses within a six years period. A general description with the most important simulator antecedents from the initial authorization for its usage to provide the ''Initial Simulator Training Course'' until currently in which it has been started the installation and testing performance of a new computer equipment that will improve and increase the simulation capacity of the Laguna Verde Nuclear Power Plant (LVNPP) simulator. In other hand, it is showed the process that the CNSNS will apply during the next verification of the simulator certification which should be performed due to the old computer equipment is being replaced with a more modern computer equipment so that the simulation capability will be improved. The verification process was discussed with the utility personnel and as result of this an agreement has been established to carry into effect this hard task. Finally, the conclusions and recommendations from regulators point of view are presented regarding to the importance of perform a well both evaluation and verification of simulators performance. (author)

Siemens Nuclear Power Corporation methods development for BWR/PWR reactor licensing

International Nuclear Information System (INIS)

Pruitt, D.W.

1992-01-01

This presentation addresses the Siemens Nuclear Power Corporation (SNP) perspective on the primary forces driving methods development in the nuclear industry. These forces are fuel design, computational environment and industry requirement evolution. The first segment of the discussion presents the SNP experience base. SNP develops, manufactures and licenses both BWR and PWR reload fuel. A review of this experience base highlights the accelerating rate at which new fuel designs are being introduced into the nuclear industry. The application of advanced BWR lattice geometries provides an example of fuel design trends. The second aspect of the presentation is the rapid evolution of the computing environment. The final subject in the presentation is the impact of industry requirements on code or methods development

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (II). The second experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-01-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the second reactor noise experiment using pseudo random binary sequences (PRBS) was performed on August 30, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 50% of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (III). The third experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-03-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the third reactor noise experiment using pseudo random binary sequences (PRBS) was performed on September 16, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 70% of reactor power and under a normal sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Tritium experience in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

1998-01-01

Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

Decommissioning Licensing Process of Nuclear Installations in Spain

International Nuclear Information System (INIS)

Correa Sainz, Cristina

2016-01-01

The Enresa experience related to the decommissioning of nuclear facilities includes the decommissioning of the Vandellos I and Jose Cabrera NPPs. The Vandellos I gas-graphite reactor was decommissioned in about five years (from 1998 to 2003) to what is known as level 2. In February 2010, the decommissioning of Jose Cabrera power plant has been initiated and it is scheduled to be finished by 2018. The decommissioning of a nuclear power plant is a complex administrative process, the procedure for changing from operation to decommissioning is established in the Spanish law. This paper summarizes the legal framework defining the strategies, the main activities and the basic roles of the various agents involved in the decommissioning of nuclear facilities in Spain. It also describes briefly the Licensing documents required to obtain the decommissioning authorization and the Enresa point of view, as licensee, on the licensing decommissioning process. (author)

Review of irradiation experiments for water reactor safety research

International Nuclear Information System (INIS)

Tobioka, Toshiaki

1977-02-01

A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

Energy Technology Data Exchange (ETDEWEB)

Gutmann, H

1972-06-05

The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

International standardization of nuclear reactor designs - the way forward

International Nuclear Information System (INIS)

Raetzke, Christian

2010-01-01

The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

77 FR 21592 - Guidelines for Preparing and Reviewing Licensing Applications for the Production of Radioisotopes

Science.gov (United States)

2012-04-10

... Licensing of Non-Power Reactors: Format and Content,'' for the production of radioisotopes and NUREG-1537, Part 2, ``Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors... by searching on http://www.regulations.gov under Docket ID NRC-2011-0135. You may submit comments by...

Safety evaluation report related to the construction permit and operating license for the research reactor at the University of Texas (Docket No. 50-602)

International Nuclear Information System (INIS)

1992-01-01

The Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC) has prepared Supplement 1 to NUREG-1135, ''Safety Evaluation Report Related to the Construction Permit and Operating License for the Research Reactor at the University of Texas'' (SER) May 1985. The reactor facility is owned by The University of Texas at Austin (UT, the applicant) and is located at the University's Balcones Research Center in Austin, Texas. This supplement to the SER (SSER) describes the changes to the reactor facility design from the description in the SER. The SER and SSER together reflect the facility as built. The SSER also documents the reviews that the NRC has completed regarding the applicant's emergency plan, security plan, and technical specifications that were identified as open in the SER

Safety Evaluation Report related to the renewal of the operating license for the TRIGA training and research reactor at the University of Arizona (Docket No. 50-113)

International Nuclear Information System (INIS)

1990-05-01

This Safety Evaluation Report for the application filed by the University of Arizona for the renewal of Operating License R-52 to continue operating its research reactor at an increased operating power level has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the University of Arizona campus in Tucson, Arizona. The staff concludes that the reactor can continue to be operated by the University of Arizona without endangering the health and safety of the public. 20 refs., 8 figs., 5 tabs

International Experience with Fast Reactor Operation & Testing

International Nuclear Information System (INIS)

Sackett, John I.; Grandy, C.

2013-01-01

Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

International exchange of safety and licensing information

International Nuclear Information System (INIS)

Lafleur, J.D. Jr.; Hauber, R.D.; Chenier, D.M.

1977-01-01

A network of formal and informal bilateral arrangements for the exchange of nuclear safety information is being established by the US Nuclear Regulatory Commission. For developing countries such arrangements can provide ready access to the extensive, fully documented safety analyses and safety research results that USNRC has accumulated. USNRC has been receiving foreign visitors at a rate of about 500 per year, largely for discussions of safety and licensing questions related to light water reactors. Exchanges also are taking place on the safety of advanced reactors. A special interest of the USNRC is in providing for reciprocal communication, at the earliest possible time, of important problems, decisions and other actions on nuclear safety matters. For example, it is essential that a newly discovered problem in a nuclear reactor be brought immediately to the attention of other governments that are responsible for the safety of similar reactors. Definite progress has been made in the USA in defining categories of information that USNRC can receive in confidence from foreign countries, and can protect from disclosure under the US Freedom of Information Act. Certain exchanges have taken place on this basis. Experience in the establishment and operation of USNRC's bilateral exchange arrangements is summarized. A typical exchange with the regulatory authority of a country building its first power reactor is described. (author)

Safety evaluation report related to the renewal of the operating license for the training and research reactor at the University of Maryland (Docket No. 50-166)

International Nuclear Information System (INIS)

1984-03-01

This Safety Evaluation Report for the application filed by the University of Maryland (UMD) for a renewal of operating license R-70 to continue to operate a training and research reactor facility has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the University of Maryland and is located at a site in College Park, Prince Georges County, Maryland. The staff concludes that this training reactor facility can continue to be operated by UMD without endangering the health and safety of the public

Advanced test reactor testing experience-past, present and future

International Nuclear Information System (INIS)

Marshall, Frances M.

2006-01-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

Introducing small modular reactors into Canada

International Nuclear Information System (INIS)

Humphries, J.R.

2012-01-01

In recent years there has been a growing interest in smaller, simpler reactors for generating electricity and process heat. This is evidenced in the growing body of literature and the increasingly frequent meetings and conferences on the subject. The interest in Small Modular Reactors (SMRs) is driven to a large extent by the desire to reduce capital costs, to reduce greenhouse gas emissions, to replace retiring fossil plants that do not meet today's environmental standards, and to provide power in locations away from large electrical grids. These drivers are as important in Canada as they are in the U.S., where the design and licensing of SMRs is being most vigorously pursued. They have led to a growing interest in Canada as a potentially significant market for SMRs, particularly in the Western Provinces of Alberta and Saskatchewan and in the remote First Nations communities of Northern Canada. There is a growing body of literature addressing the regulation and licensing of Small Modular Reactors in the U.S. Issues being identified in there can generally be categorized as licensing framework issues, licensing application issues, and design and manufacturing issues. Many of these issues are embedded in the US regulatory framework and can only be resolved through changes in the regulations. For the most part these issues are equally applicable in Canada and will need to be addressed in introducing SMRs here. A significant difference, however, is that these issues can be addressed within the Canadian regulatory framework without requiring changes in the regulations. The CNSC has taken a very proactive stance regarding the licensing of small reactors in Canada. They have published two new Regulatory Documents stipulating the requirements for licensing small reactors. A key feature is that they allow the application of a 'graded approach' in which the stringency of the design measures and analyses applied are commensurate with the level of risk posed by

Interim licensing criteria for physical protection of certain storage of spent fuel

International Nuclear Information System (INIS)

Dwyer, P.A.

1994-11-01

This document presents interim criteria to be used in the physical protection licensing of certain spent fuel storage installations. Installations that will be reviewed under this criteria are those that store power reactor spent fuel at decommissioned power reactor sites; independent spent fuel storage installations located outside of the owner controlled area of operating nuclear power reactors; monitored retrievable storage installations owned by the Department of Energy, designed and constructed specifically for the storage, of spent fuel; the proposed geologic repository operations area; or permanently shutdown power reactors still holding a Part 50 license. This criteria applies to both dry cask and pool storage. However, the criteria in this document does not apply to the storage of spent fuel within the owner-controlled area of operating nuclear power reactors

Channel type reactors with supercritical water coolant. Russian experience

International Nuclear Information System (INIS)

Kuznetsov, Y.N.; Gabaraev, B.A.

2003-01-01

Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

License renewal in the United States

International Nuclear Information System (INIS)

Brons, Jack

2002-01-01

Full text: Nuclear plants in the United States are licensed for 40 years, a length specified in the Atomic Energy Act of 1954, which laid out much of the regulatory basis for the commercial nuclear industry. The Act, however, made provision for license renewal. The original 40-year license period was chosen arbitrarily by the U.S. Congress because it was the typical period over which utilities recovered their investment in electricity generating plants. Nuclear plants, however, are subject to a rigorous program of Nuclear Regulatory Commission oversight, maintenance and equipment replacement. In effect, they must be in the same operating condition on the last day of their licenses as they were on the first. As the industry matured, it became apparent that there was no physical limitation on the continued operation of nuclear plants past 40 years. The industry turned its attention toward license renewal. When the issue was first raised, the NRC considered stringent process equivalent to seeking a new operating license for each plant. The complexity, length and cost of the process made it unlikely that many nuclear plants would seek license renewal. The nuclear industry worked successfully with NRC on the application of generic principles to license renewal, however, and in 1995, the NRC issued an efficient, tightly-focused rule that made license renewal a safe, viable option. To extend the operating license for a reactor, a company must demonstrate to the NRC that aging effects will be adequately managed during the renewal terms, thus ensuring equipment functionality. The rule allows licensees to apply for extensions of up to 20 years. The first license renewal application was filed in 1998 by the owner of the two-unit Calvert Cliffs plant. Shortly thereafter, an application was filed for the three-unit Oconee Nuclear Station. The NRC renewed the licenses for all five units in 2000, and since then, five more licenses have been renewed. The NRC has received 37

Detecting Dark Photons with Reactor Neutrino Experiments

Science.gov (United States)

Park, H. K.

2017-08-01

We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ reactors as potential sources of intense fluxes of low-mass dark photons.

FRM-II: status of construction, licensing fuel tests

International Nuclear Information System (INIS)

Axmann, A.; Boening, K.; Nuding, M.; Didier, H.J.

2001-01-01

The research reactor FRM-II of the Technische Universitaet Muenchen is now ready for the nuclear start-up, but still waiting for the operational license. The high-flux neutron-source FRM-II (8 x 10 14 n/(s cm 2 ) is a unique tool for solid state physics and materials research by neutron scattering, positron annihilation experiments and activation analysis, as well as for fundamental physics, isotope production, silicone doping, cancer therapy by irradiation with fission neutrons and for tomography with fast and thermal neutrons. Reactor built in facilities as a hot source, a cold source, an uranium loaded converter plate producing an intense beam of fission neutrons, allow to expand the range of usable neutron energies far beyond the thermal spectrum. In addition, a source providing an intense beam of fission products is planned to be constructed by the Ludwig-Maximilians-Universitaet Muenchen and a source of ultra cold neutrons is planned by the Physics Department of the Technische Universitaet Muenchen. The reactor is already prepared for both of these facilities. (author)

Considerations for advanced reactor design based on EBR-II experience

International Nuclear Information System (INIS)

King, R. W.

1999-01-01

The long-term success of the Experimental Breeder Reactor-II (EBR-II) provides several insights into fundamental characteristics and design features of a nuclear generating station that enhance safety, operability, and maintainability. Some of these same characteristics, together with other features, offer the potential for operational lifetimes well beyond the current licensing time frame, and improved reliability that could potentially reduce amortized capital costs as well as overall operation and maintenance costs if incorporated into advanced plant designs. These features and characteristics are described and the associated benefits are discussed

Review on conformance of JMTR reactor facility to safety design examination guides for water-cooled reactors for test and research

International Nuclear Information System (INIS)

Ide, Hiroshi; Naka, Michihiro; Sakuta, Yoshiyuki; Hori, Naohiko; Matsui, Yoshinori; Miyazawa, Masataka

2009-03-01

The safety design examination guides for water-cooled reactors for test and research are formulated as fundamental judgements on the basic design validity for licensing from a viewpoint of the safety. Taking the refurbishment opportunity of the JMTR, the conformance of the JMTR reactor facility to current safety design examination guides was reviewed with licensing documents, annexes and related documents. As a result, it was found that licensing documents fully satisfied the requirements of the current guides. Moreover, it was found that the JMTR reactor facility itself also satisfied the guides requirements as well as the safety performance, since the facility with safety function such as structure, systems, devices had been installed based on the licensing documents under the permission by the regulation authority. Important devices for safety have been produced under authorization of regulating authority. Therefore, it was confirmed that the licensing was conformed to guides, and that the JMTR has enough performance. (author)

Decay power evaluation for licensing analysis

International Nuclear Information System (INIS)

Tran, H.; Schrock, V.E.

1987-01-01

The ANSI/ANS 5.1-1979 Standard on Decay Power in shutdown reactors has been available as the basis for accident analysis for the past 7 yr. The US Nuclear Regulatory Commission has made a commitment to use this standard in new licensing approaches and has approved a licensing model for boiling water. More sweeping changes in the licensing rules are currently under review that will involve the use of best-estimate models and a statistical evaluation of the uncertainty (95% confidence level) in the key results. The structure of the decay power standard is well suited for such applications because it provides a statistically meaningful uncertainty in the decay power from fission products. The normalized decay power is a function specific to each point in the reactor volume due to the fact that the fuel composition develops a spatial dependence as burnup proceeds and decay power depends on the mix of fissioning nuclides. For reactor safety calculations it is desirable to employ a single temporal decay power function for the whole core inasmuch as many variations of accident parameters are required. This is the usual approach in large system thermal-hydraulics codes. Such a single representative or generic curve for a specified total operating power history can be acceptable but at the expense of some increase in the uncertainty. In this paper, the author present a method of evaluating the additional uncertainty in the decay power associated with use of a generic curve

Storage experience in Hungary with fuel from research reactors

International Nuclear Information System (INIS)

Gado, J.; Hargitai, T.

1996-01-01

In Hungary several critical assemblies, a training reactor and a research reactor have been in operation. The fuel used in the research and training reactors are of Soviet origin. Though spent fuel storage experience is fairly good, medium and long term storage solutions are needed. (author)

Investigating the spectral anomaly with different reactor antineutrino experiments

Directory of Open Access Journals (Sweden)

C. Buck

2017-02-01

Full Text Available The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in U235 with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted U235 spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

Operational experience of the Marcoule reactors

International Nuclear Information System (INIS)

Conte, F.

1963-01-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

Light-water reactor accident classification

International Nuclear Information System (INIS)

Washburn, B.W.

1980-02-01

The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

MAPLE research reactor safety uncertainty assessment methodology

International Nuclear Information System (INIS)

Sills, H.E.; Duffey, R.B.; Andres, T.H.

1999-01-01

The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

Future of Nuclear Power: NRC emergency preparedness licensing activities agenda

International Nuclear Information System (INIS)

Essig, T.H.

1995-01-01

This talk summary addresses the issue of how future policies of the NRC will affect nuclear power in areas such as construction, emergency preparedness, and licensing. Specific topics covered include the following: Emergent EP licensing issues for operating nuclear Power Plants; 10CFR Part 52 and the process for licensing of Advanced Light Water Reactors (ALWRs); and potential revisions to emergency preparedness programs for future nuclear power plants

Nuclear Reactor Sharing Program

International Nuclear Information System (INIS)

1994-01-01

The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

Leak before break experience in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Moan, G.D.; Coleman, C.E.

1988-04-01

The paper describes how the requirements for Leak-Before-Break are met in CANDU reactors. The requirements are based on operational and laboratory experience. After the onset of leakage in a fuel channel from a delayed hydride crack, time is available to the operator to take action before the crack grows to an unstable length. The time available is calculated using different models which use crack growth data from small specimen tests. When the results from crack growth behaviour experiments, carried out on components removed from reactor are used in the model, the time available for operator response is about 100 hours

U.S. NRC training for research and training reactor inspectors

International Nuclear Information System (INIS)

Sandquist, G.M.; Kunze, J.F.

2011-01-01

Currently, a large number of license activities (Early Site Permits, Combined Operating License, reactor certifications, etc.), are pending for review before the United States Nuclear Regulatory Commission (US NRC). Much of the senior staff at the NRC is now committed to these review and licensing actions. To address this additional workload, the NRC has recruited a large number of new Regulatory Staff for dealing with these and other regulatory actions such as the US Fleet of Research and Test Reactors (RTRs). These reactors pose unusual demands on Regulatory Staff since the US Fleet of RTRs, although few (32 Licensed RTRs as of 2010), they represent a broad range of reactor types, operations, and research and training aspects that nuclear reactor power plants (such as the 104 LWRs) do not pose. The US NRC must inspect and regulate all these entities. This paper addresses selected training topics and regulatory activities provided US NRC Inspectors for US RTRs. (author)

Fast breeder reactors: can we learn from experience

International Nuclear Information System (INIS)

Keck, O.

1981-01-01

An economic analysis of FBRs, in particular the long-term benefits to be expected, with reference to the experience of the West German fast breeder reactor programme suggests ways of bringing more realism into governmental decisions on the development of new reactor types. It is suggested that if reactor manufacturers and utilities financed commercial-size demonstration plants from their own funds, then the government would get more realistic advice. (U.K.)

Safety-evaluation report related to the renewal of the operating license for the research reactor at the Iowa State University (Docket No. 50-116)

International Nuclear Information System (INIS)

1983-09-01

This Safety Evaluation Report for the application filed by the Iowa State University (ISU) for a renewal of the Class 104 Operating License R-59 to continue to operate its Argonaut-type research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Iowa State University, and is located on the ISU campus in Ames, Story County, Iowa. The staff concludes that the reactor facility can continue to be operated by ISU without endangering the health and safety of the public. The principal matters reviewed are: design, testing, and performance of the reactor components and systems; the expected consequences of credible accidents; the licensee's management organization; the method used for the control of radiological effluents; the licensee's technical specifications; financial data and information; the physical protection program; procedures for training reactor operators; and emergency plans. 11 references, 15 figures, 13 tables

Operating US power reactors

International Nuclear Information System (INIS)

Silver, E.G.

1988-01-01

This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

International Nuclear Information System (INIS)

Server, W.L.; Nanstad, R.K.; Odette, G.R.

2012-01-01

The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

Licensed operating reactors. Status summary report data as of 12-31-94: Volume 19

International Nuclear Information System (INIS)

1995-04-01

The Nuclear Regulatory Commission's annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar year 1994) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

Licensed operating reactors. Status summary report data as of 12-31-94: Volume 19

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-04-01

The Nuclear Regulatory Commission`s annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar year 1994) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided.

Review of the first partial licence for the Muelheim-Kaerlich reactor on the basis of the licensing requirements under nuclear law

International Nuclear Information System (INIS)

Hahn, L.; Wimmer, J.

1990-11-01

The expert opinion is restricted in the main to the licensing requirements in connection with the latest state-of-the-art precautionary measures against damage, i.e. subsection 3 of paragraph 7 section 2 Atomic Energy Act. Subject of the evaluations regarding the required new scope of licensing were the cancelled first partial licence, subsequent partial licences, the TUEV (Technical Control Board) expert opinion, the safety report with its continuations as well as the judgments of the Higher Administrative Court in Koblenz and the Federal Administrative Court. For determining the state-of-the-art, the expert opinion refers to the valid body of regulations, the latest concepts for PWRs in particular in the Federal Republic of Germany, and more recent results in reactor safety research. (orig./DG) [de

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-09-01

The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

Safety Evaluation Report related to the renewal of the operating license for the TRIGA training and research reactor at the University of Utah (Docket No. 50-407)

International Nuclear Information System (INIS)

1985-03-01

This Safety Evaluation Report for the application filed by the University of Utah (UU) for a renewal of operating license R-126 to continue to operate a training and research reactor facility has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the University of Utah and is located on its campus in Salt Lake City, Salt Lake County, Utah. The staff concludes that this training reactor facility can continue to be operated by UU without endangering the health and safety of the public

License renewal and power upgrade of the Cornell University TRIGA reactor

International Nuclear Information System (INIS)

Aderhold, Howard C.

1984-01-01

The Cornell Mark II TRIGA reactor has been a principal facility for instruction and research in nuclear science and engineering at Cornell, and it has been extensively used by other departments at Cornell and by nearby universities and industries. Initially the fuel was low hydride, 8.5w/o 19%-enriched, aluminum clad; in 1974 it was changed to high-hydride, stainless-steel-clad. The maximum power has been 100 kW, with pulses to $2, and operation has been on a one-shift demand basis. Annual energy generation of 50 MWH has been typical. Standard features include a 4-inch tangential port and our 6-inch radial ports, a thermal column with hohlraum and vertical access, a central thimble, a 'rabbit', and a set of dry irradiation tubes, replacing the 'Lazy Susan'. The license was renewed and amended in November 1983; the new limits are 500 kW and $3 pulses. Physical changes to the facility included addition of a water-to-water heat exchanger and of a diffuser at the water outlet ∼ 60 cm above the core. The flow rate is 300 liters per minute in the primary (reactor) side of the heat exchanger. The temperature of the chilled water entering the secondary of the exchanger is ∼ 12?C; its flow rate is adjusted by a servo-controlled by-pass valve to maintain the desired range of pool water temperature. Steps taken to go to higher power included rearrangement of fuel elements to increase excess reactivity, recalibration of control rods, and power vs ion chamber current calibrations at successively higher power by comparing the rate of rise of pool temperature with a known rate using electrical heating elements. Steady-state operation has been done up to 480 kW (nominal) but pulsing at the newly allowed higher levels has not been tested as yet

77 FR 70837 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Science.gov (United States)

2012-11-27

... intervene shall be filed in accordance with the Commission's ``Rules of Practice for Domestic Licensing... request: The proposed change will delete Function 14, SG [Steam Generator] Water Level--Low, Coincident... capability to shut down the reactor when required on Low--Low Steam Generator water level. The ability to...

Radiological characterisation experience with Magnox reactors

International Nuclear Information System (INIS)

Westall, Bill A.; Towton, Barbara

2012-01-01

At the end of generation, power reactors will be decommissioned. Whether decommissioning is prompt or deferred, knowledge of the radioactive inventory of plant and structures is needed to develop and underpin the decommissioning strategy. As decommissioning progresses the level of detail required for the radioactive inventory increases as more specific and detailed questions need answering. Failure to adequately characterise will result in increased costs and project overruns due to missing optimal solutions, over pessimistic assumptions or unforeseen problems and regulatory issues. Radiological characterisation for decommissioning of Magnox power stations in the UK has been in progress for over a quarter of a century. Firstly measurements and calculations were carried out to develop a strategy. These have been followed by measurements to determine radioactive inventories of waste streams and packages or to allow decontamination of structures and most recently for partial de-licensing of sites. Some examples of the work carried out for the Magnox stations will be given, ranging from the neutron activation calculations to estimate the radioactive inventory within a bio-shield and measurements to validate them. Various plant and structures where the radioactive inventory is due to contamination have been characterised by measurements and examples for boilers and cooling ponds will be discussed. Various routine and ad-hoc measurements and shielding assessments have been performed on waste forms to help satisfy conditions for acceptance for disposal or exemption, which will be reviewed. Finally the measurements for de-licensing and the successful application of Data Quality Objectives will be addressed. (authors)

Inspection of licensed nuclear power plants in the United States

International Nuclear Information System (INIS)

Thornburg, H. D.

1977-01-01

Inspection of licensed nuclear power plants in the United States is performed by the Office of Inspection and Enforcement (IE), United States Nuclear Regulatory Commission. IE has several key functions : a) Inspection of licensees and investigation of incidents, occurrences and allegations. b) Detection and correction of safety and security problems. c) Enforcement of rules, regulations, and Commission orders. d) Feedback to the industry and others regarding safety experience. e) Informing the public and others. Major enforcement actions and events involving operating power reactors for the past several years will be summarized. (author)

Advanced reactor safety research quarterly report, October-December 1982. Volume 24

Energy Technology Data Exchange (ETDEWEB)

None

1984-04-01

This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-04-01

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

Libraries and licensing

Directory of Open Access Journals (Sweden)

Maja Žumer

2001-01-01

Full Text Available In the mid 90s, the abundance of various electronic publications exposed libraries to the problems of licensing electronic content. Various licensing principles have been prepared recently to help libraries in the process; it can be said that in general, the knowledge of licensing issues has improved in libraries of all types. Libraries form consortia in order to gain stronger negotiating positions and obtain better conditions.In the article, new licensing principles are presented in more detail, as well as some domestic and foreign experiences with consortia forming.

Initiatives in transport cask licensing

International Nuclear Information System (INIS)

Patterson, John

1998-01-01

The variations in research reactor fuel form, configuration, irradiation characteristics, and transport cask have required a substantial number of transport cask licensing actions associated with foreign research reactor spent fuel transportation. When compounded by limited time for shipment preparations, due to contract timing or delayed receipt of technical data, the number and timing of certifications has adversely impacted the ability of regulatory agencies to support intended shipping schedules. This issue was brought into focus at a april, 1998 meeting among DOE, the US Nuclear Regulatory Commission, and DOE's spent fuel transportation contractors. (author)

Licensed operating reactors: Status summary report, Data as of 12-31-86

International Nuclear Information System (INIS)

1987-09-01

This report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are monthly highlights and statistics for commercial operating units and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

Proposed Activities to Address Regulatory Gaps and Challenges for Licensing Advanced Reactors Using Seismic Isolation

International Nuclear Information System (INIS)

Coleman, Justin Leigh; Kammerer, Annie M.; Whittaker, Andrew S.

2016-01-01

Over the last decade, particularly since implementation of the certified design regulatory approaches outlined in 10 CFR 52, 'Licenses, Certifications, and Approvals for Nuclear Power Plants,' interest has been increasing in the use of seismic isolation (SI) technology to support seismic safety in nuclear facilities. In 2009, the United States (U.S.) Nuclear Regulatory Commission (NRC) initiated research activities to develop new guidance targeted at isolated facilities because SI is being considered for nuclear power plants in the U.S. One product of that research, which was developed around a risk-informed regulatory approach, is a draft NRC NUREG series (NUREG/CR) report that investigates and discusses considerations for use of SI in otherwise traditionally founded large light water reactors (LWRs). A coordinated effort led to new provisions for SI of LWRs in the American Society of Civil Engineers standard ASCE/SEI 4-16, 'Seismic Analysis of Safety Related Nuclear Structures.' The risk-informed design philosophy that underpinned development of the technical basis for these documents led to a set of proposed performance objectives and acceptance criteria intended to serve as the foundation for future NRC guidance on the use of SI and related technology. Although the guidance provided in the draft SI NUREG/CR report and ASCE/SEI 4 16 provides a sound basis for further development of nuclear power plant designs incorporating SI, these initial documents were focused on surface-founded or near-surface-founded LWRs and were, necessarily, limited in scope. For example, there is limited information in both the draft NUREG/CR report and ASCE/SEI 4-16 related to nonlinear analysis of soil-structure systems for deeply-embedded reactors, the isolation of components, and the use of vertical isolation systems. Also not included in the draft SI NUREG/CR report are special considerations for licensing of isolated facilities using the certified design

Proposed Activities to Address Regulatory Gaps and Challenges for Licensing Advanced Reactors Using Seismic Isolation

Energy Technology Data Exchange (ETDEWEB)

Coleman, Justin Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kammerer, Annie M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Whittaker, Andrew S. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-12-01

Over the last decade, particularly since implementation of the certified design regulatory approaches outlined in 10 CFR 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” interest has been increasing in the use of seismic isolation (SI) technology to support seismic safety in nuclear facilities. In 2009, the United States (U.S.) Nuclear Regulatory Commission (NRC) initiated research activities to develop new guidance targeted at isolated facilities because SI is being considered for nuclear power plants in the U.S. One product of that research, which was developed around a risk-informed regulatory approach, is a draft NRC NUREG series (NUREG/CR) report that investigates and discusses considerations for use of SI in otherwise traditionally founded large light water reactors (LWRs). A coordinated effort led to new provisions for SI of LWRs in the American Society of Civil Engineers standard ASCE/SEI 4-16, “Seismic Analysis of Safety Related Nuclear Structures.” The risk-informed design philosophy that underpinned development of the technical basis for these documents led to a set of proposed performance objectives and acceptance criteria intended to serve as the foundation for future NRC guidance on the use of SI and related technology. Although the guidance provided in the draft SI NUREG/CR report and ASCE/SEI 4 16 provides a sound basis for further development of nuclear power plant designs incorporating SI, these initial documents were focused on surface-founded or near-surface-founded LWRs and were, necessarily, limited in scope. For example, there is limited information in both the draft NUREG/CR report and ASCE/SEI 4-16 related to nonlinear analysis of soil-structure systems for deeply-embedded reactors, the isolation of components, and the use of vertical isolation systems. Also not included in the draft SI NUREG/CR report are special considerations for licensing of isolated facilities using the certified design approach in 10 CFR

Safety and licensing issues for Indian PHWRs

International Nuclear Information System (INIS)

Srinivasan, G.R.; Das, M.

1997-01-01

India has achieved competency in design, construction, commissioning and operation of Pressurized Heavy Water Reactor based Nuclear Power Plants and has completed more than 120 reactor operating years with an extremely satisfactory safety record. In this paper, the safety management in NPCIL and operational safety aspects are discussed, licensing and regulatory approach is described and some of the main safety issues for Indian PHWRs are brought out. (author)

Development of quantitative goals for inherent safety feature design and licensing

International Nuclear Information System (INIS)

Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.; Okrent, D.

1987-01-01

There is now considerable interest in the development of advanced fast reactors whose major focus is inherent safety. The achievement of inherent safety can be viewed from several aspects. In the Integral Fast Reactor Concept the approach is to utilize the intrinsic characteristics of pool-type liquid metal fast breeder reactors (LMFBRs) and the properties of metal fuels to integrate a high degree of inherent safety into the design. The PRISM and SAFR concepts focus on other inherent safety features. The reactors discussed above represent a radical departure from existing LWR designs as well as previous LMFBR designs (e.g., CRBRP) which are based, for the most part, on the General Design Criteria found in 10CFR50 Appendix. In view of these parallel developments (advanced reactors exploiting inherent safety and the use of quantitative goals to augment licensing), there appears to be a need to perform research on the development of methods for designing, assessing, and licensing inherent safety features in advanced reactors. The objectives of such research are outlined

E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

Energy Technology Data Exchange (ETDEWEB)

Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

2015-08-15

Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

Review of experiments for research reactors - approved 1974

International Nuclear Information System (INIS)

Anon.

1977-01-01

This standard establishes guidelines for the review and approval of experiments performed at research reactor facilities. This standard identifies the major areas that shall be reviewed for each experiment to ensure that it (a) falls within the limits delineated in the technical specifications, (b) does not present an unreviewed safety question as defined in 10 CFR Section 50.59 π2-, (c) does not constitute a threat to the health and safety of any individual or group of individuals, and (d) does not constitute a hazard to the reactor facility or other equipment. In addition, this standard recommends a system for classifying experiments to establish levels of review and approval commensurate with the level of risk inherent in the experiment

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-01-01

This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

Atomic Safety and Licensing Board Panel annual report, fiscal year 1989

International Nuclear Information System (INIS)

Cotter, B.P. Jr.

1990-07-01

In Fiscal Year 1989, the Atomic Safety and Licensing Board Panel (ASLBP) handled 40 proceedings involving the construction, operation and maintenance of commercial nuclear power reactors or other activities requiring a license from the Nuclear Regulatory Commission. This report summarizes, highlights and analyzes how the wide-ranging issues raised in these proceedings were addressed by the Judges and Licensing Boards of the ASLBP during the year. 5 figs., 3 tabs

The problem of licensing and safety of nuclear power plants

International Nuclear Information System (INIS)

Silva, R.A. da.

1987-01-01

The historical evolution of licensing process of nuclear power plants is presented. The designs carried out by FURNAS for constructing Angra-1 reactor and its contribution to the Brazilian CNEN in de licensing process, are evaluated. The aims of FURNAS Research Programs are determined and the safety goals are established. (M.C.K.) [pt

Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-04-01

This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee`s annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs.

Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

International Nuclear Information System (INIS)

1997-04-01

This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee's annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

International Nuclear Information System (INIS)

Bess, John D.; Gullifor, Jim

2015-03-01

The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

Safety evaluation report related to the renewal of the operating license for the Research Reactor at the State University of New York at Buffalo, Docket No. 50-57

International Nuclear Information System (INIS)

1983-05-01

This Safety Evaluation Report for the application filed by the State University of New York at Buffalo for a renewal of Operating License R-77 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned by the State University of New York and is located on the campus in Buffalo, New York. Based on its technical review, the staff concludes that the reactor facility can continue to be operated by the University without endangering the health and safety of the public or endangering the environment

Research reactors. Problems of fuel element enrichment reduction. Deliberations and comments

International Nuclear Information System (INIS)

1978-10-01

This paper summarises the main data from the major research reactors in the Federal Republic of Germany utilising highly enriched uranium (HEU) and presently available fuel technology for their fuel elements. The required modification for an adaption of the fabrication to lower enriched fuel are considered as well as the consequences on reactor performance operation and licensing. On the basis of past experience with reactor modifications a rough estimate of 82 months is given for the conversion of a reactor to a modified type of fuel and of 70 months for a fuel test program. The conclusions reflect the own calculations and data from other papers submitted to INFCE-WG 8C

Semi-annual report of the licensing program - July to December 1988

International Nuclear Information System (INIS)

Anon.

1989-01-01

The main activities developed and described are: licensing of CIPC (FSAR), emergency plan of NUCLEI, reevaluation of the maximum credible accident of NUCLEI, updating of the safety analysis report of CDTN, emergency plan of IPR-R1 reactor and licensing of the DIENR.CN. installations. (E.G.) [pt

Experience and prospects for developing research reactors of different types

International Nuclear Information System (INIS)

Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

2015-01-01

NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

Technical report on operating experience with boiling water reactor offgas systems

International Nuclear Information System (INIS)

Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

1978-03-01

Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-01

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems