bluuebunny/crossref_metadata_2025_split huggingface dataset:
Risk-Informed Structural Integrity Management: Development of SMR-Centric Piping Reliability Models
Authors: Lydell Bengt 
Published: July 2014

In the context of risk-informed applications, this paper addresses the progress with piping reliability analysis methods and techniques and their role in supporting development of risk-informed structural integrity programs for small modular reactors (SMRs). The structural integrity of a pressure boundary is determined by multiple and interrelated reliability attributes and influence factors. Depending on the conjoint requirements for damage and degradation, certain combinations of material, operating environment, loading conditions together with applicable design codes and standards, certain passive components are substantially more resistant to damage and degradation than others. As an example, for stabilized austenitic stainless steel pressure boundary components, there are no recorded events involving active, through-wall leakage. By contrast, for unstabilized austenitic stainless steel, multiple events involving through-wall leakage have been recorded, albeit with relative minor leak rates.
The field experience with safety- and non-safety related piping in commercial GenI through GenIII nuclear power reactors is quite extensive. Equally extensive is the experience gained from the implementation of different degradation mechanism mitigation strategies. By applying advanced piping reliability models, this body engineering data and integrity management insights can be used to assess the projected structural integrity of new piping system designs, including those of SMRs. The paper presents an overview of recent methodological advances and insights from the application of statistical piping reliability models to advanced reactor designs. Examples are provided on how piping reliability parameter estimates are affected by different integrity management strategies as well as by advanced, degradation mechanism (DM) resistant materials. The technical basis for the work that is presented in this paper has evolved over a period of 20+ years of focused and sustained R&D in the area of statistical models of piping reliability.
Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)
Authors: Lippy Ronald C. 
Published: July 2014

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs).
The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.
ASME Codes and Small Modular Reactors
Authors: Demoss Dennis , Janusz Stefan J. , Niemer Richard P. 
Published: July 2014

Authors: Sargent & Lundy, LLC: Dennis Demoss, Stefan Janusz, and Richard Niemer. Focus: Codes and Standards. This paper will review new Small Modular Reactor (SMR) designs and implications on ASME Codes and Standards. SMR technologies which will be discussed include B&W mPower (TVA Clinch River site), Westinghouse SMR (Ameren Callaway site), NuScale/Fluor SMR (Pacific NW lab site), Holtec SMR (Savannah River site), and Non-U.S. Designs such as the Russian SMRs.
U.S. SMR reactor designs vary from 45 to approximately 225 MWe. SMRs, as defined by the International Atomic Energy Agency, have an electrical output less than 700 MW [1]; however, U.S. SMRs are typically defined as producing less than 350 MWe. SMR goals include significantly reducing plant capital cost requirements and enabling multi-reactor module construction and addition over time providing greater utility flexibility. A primary SMR advantage includes its installation in smaller grids typical of electrical power systems in developing countries. Unique aspects of the SMR technologies include integral reactor and steam generator vessel, integral pressurizer and internal piping, below-grade containment vessel, helical-coil integral steam generators, integral decay heat removal systems, modular plant construction and arrangement, in-service inspection unique requirements, special materials, and welding. SMR technologies provide unique challenges for conformance with ASME Codes and Standards.
Non-U.S. designs, such as the Russian floating ship-type configuration SMR (KLT-40S) and Russian land-based SMR (VBER-300), will be discussed from the perspective of compatibility with ASME Codes. Discussion will be provided regarding non-U.S. SMR operational safety and inspection requirements based on ASME Codes.
SMR development may require the expansion and clarification of current ASME Code design rules and requirements. For example, Section XI testing requirements and frequencies may require revision due to longer intervals between refueling. New divisions of ASME Codes may be required to address inspection of SMR inaccessible plant components and materials and plant operational differences from previous LWR designs. Additional reactor internal components and equipment may also require additional ASME Code considerations. Finally, higher design temperatures resulting from passive design shutdown considerations may require development and Code acceptance of new materials.
Issues of Insurance of Civil Liability for Nuclear Damage From Nuclear Low Power Plants
Authors: Demin Vladimir F. , Kuznetsov Vyacheslav P. 
Published: July 2014

In the frame of economic analysis of nuclear power (NP) with SMRs in their total life cycle the expert analysis of the problems associated with the civil liability for nuclear damage from SMRs on the example of transportable nuclear power units (TNUs) was performed.
Purpose of the analysis is as follows:
• Assessment of NP’ safety and economy changes in its development based on TNUs with KLT-40S and partly with RITM-200M reactor units.
• Work out of recommendations on this development’ direction in terms of the insurance approach justification and amount of compensation for nuclear damage.
The following aspects were considered in the analysis:
1. National and international approaches and practice of nuclear insurance.
2. Specific features of TNUs and differences from large NPPs basing on example of the design of floating power unit FPU “Academician Lomonosov” with KLT-40S reactors.
3. Assessment of severe accident consequences during TNU’ life cycle.
4. Analysis of insurances’ approaches and assessment of possible insurance costs.
Nuclear Power System on a Base of Transportable Nuclear Power Plants
Authors: Velikhov Evgeny P. , Kuznetsov Vyacheslav P. , Demin Vladimir F. 
Published: July 2014

This paper presents the initial provisions, materials, results, current status and next tasks of the study dedicated to the issues of legal and institutional support of transportable nuclear power plants. This study is performed in the framework of the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles INPRO.
Transportable nuclear power plants (TNPPs) are either small nuclear power plants (SNPPs) with their lifecycle implemented on a single transportable platform, or SNPPs assembled of transportable factory-made modules.
Advantages of SNPPs and TNPPs are:
• Enhanced safety and reliability;
• Design simplicity,
• Shorter construction period;
• Industrial serial production;
• Smaller capital costs and shorter investment cycle compared with large NPP;
• Possibility of autonomous operation;
• Suitability for non-electric application and others.
There is an objective evidence of growing interest in developing a nuclear energy system (NES) based on SNPPs including TNPPs.
Underlying assumptions of the Russian study:
• The User of TNPP services is interested in receiving energy only, does not claim ownership of nuclear technologies, materials and TNPP itself, and this incurs minimal liability for nuclear energy use; INPRO defines this TNPP lifecycle option as “Maximum outsourcing”;
• All operations involving nuclear fuel are performed either at the TNPP manufacturer plant, or at a regional TNPP service center within the Holder’s liability zone;
• TNPP sitting requires no onsite operations except assembling.
Expert reviews have been performed to confirm TNPP lifecycle compliance with the nuclear legislation in fields such as: safety; non-proliferation; nuclear materials’ monitoring, accounting and control; physical protection; and civil liability for nuclear damage; transport operations.
It was confirmed that:
• In traditional approaches, the existing legal and institutional framework is sufficient for implementing TNPP lifecycle; to achieve the highest efficiency and safety of TNPPs it is necessary to develop TNPPs’ designs, their legal and institutional support;
• The following issues are of immediate interest for further studies: combination of inherent safety features and passive safety systems in TNPPs; TNPP lifecycle economy; lifecycle concept without onsite refueling; new approaches to indemnification for nuclear damage; new approaches to physical protection; nuclear liability of TNPP User; remote nuclear materials monitoring, and control and TNPP’ operating; serial industrial fabrication; licensing and certification; public-private partnership; international personnel training system; international cooperation in TNPP fabrication and servicing; role of the IAEA in developing TNPP-based NES.
• TNPP/SNPP-based nuclear energy system including all kinds of respective legal, institutional and infrastructural support should become the subject of further studies.
Incorporation of Fukushima Lessons Learned Into Small Modular Light Water Reactor Functional Requirements
Authors: Loflin Leonard I. 
Published: July 2014

This paper describes the EPRI (Electric Power Research Institute) project to enable the incorporation of Fukushima Lessons Learned into new Advanced Light Water Reactor (ALWR) nuclear plant designs and projects. This includes large scale plants and Small Modular Reactors (SMR).
EPRI, working with the bulk of the US nuclear industry, compiled the Utility Requirements Document (URD) for new passive and evolutionary ALWRs. The ∼3,400 pages of functional design and project requirements in the URD became one of the bases documents for the current AP1000, ESBWR and EPR designs. The URD was and is a declaration of requirements of Owners/Operators for any new nuclear plant they might purchase. The 2013 revision 12 of the URD incorporated specific design and project requirements that are directly linked to industry summary conclusions and recommendations concerning the Fukushima event. During 2013 and 2014, EPRI, with assistance from the Department of Energy (DOE), is developing the URD requirements for selected SMRs.
A Retrospective Look on Plant Events for Evolving Expectations in Nuclear Safety
Authors: Koshy Thomas 
Published: July 2014

The nuclear industry continues to rise above challenges resulting from major plant events around the world. It is important to study the significant events, and find solutions to overcome the vulnerabilities identified, and retain the lessons while technology evolves to the next generation of SMRs. The plant events discussed are from Light Water reactors (LWRs) but he derived lessons could apply to LWR–SMRs and it could be extrapolated for other designs.
The historical lessons on Station-Black-Out needs to be re-examined further in a new dimension in the light of “Fukushima type” events where AC & DC power was lost to the entire station from a common cause and the recovery in a reasonable period was not practical.
The SMRs are specially considered for smaller grid size and for remote locations where a large reactor is unsuitable. The specific advantages for SMRs are its smaller foot print, modular type construction and flexibility for incremental units to meet rising demand in power. These features are very attractive to countries with small grids and for countries that are in the early stages of nuclear power technology development. In anticipation of these opportunities, when we launch into new technology of evolutionary and revolutionary advancements, the reactors have to be more rugged with ample diversity in critical reactor safety areas through both passive and active means to account for a conditions that may inhibit electrical energy as a source of motive power.
An electrical event in Forsmark plant that propagated from an electrical switchyard resulted in two core cooling divisions disabled and consequently exacerbating the plant condition by opening the steam relief system for reactor coolant system and that significantly increased the probability for core damage. A minor spark in an electronic control system card in a US plant caused inadvertent actuation of emergency core cooling and disabled the capability of control room operators to intervene and prevent the primary cooling system from completely going water solid in the primary loop of Reactor Coolant System. A renewed assessment is needed to address the following questions for advancing reactor safety in the new evolving generation of LWR-SMR plants and to other types as applicable to preserve the historic lessons in safety.
(1) Evaluate the diversity in emergency core cooling systems following loss of all AC power onsite;
(2) Ensure diversity in reactor trip & depressurization;
(3) Evaluate the suitability of logic/control system failure mode resulting from power supply failures in instrument channels and/or divisions, i.e. by conducting failure mode and effects analysis (FMEA) for system, power supplies and component level;
(4) Confirm independence in reactor trip and emergency core cooling systems for sensors, power supplies and actuation systems.
Complete Genome Sequences of Three Iberian Brucella suis Biovar 2 Strains Isolated from Wild Boars
Authors: Ferreira Ana Cristina , Tenreiro Rogério , Corrêa de Sá Maria Inácia , Dias Ricardo 
Published: July 2014

ABSTRACT

Brucella suis
biovar 2 is the most common biovar isolated from wild boars (
Sus scrofa
) associated with transmission to outdoor-reared pigs in Europe. We report here the complete and annotated genome sequences of three strains isolated from wild boars in Portugal and Spain and belonging to the Iberian clone (haplotypes 2d and 2e).
Complete Mitochondrial Genome Sequence of Lichtheimia ramosa (syn.
Lichtheimia hongkongensis
)
Authors: Leung Shui-Yee , Huang Yi , Lau Susanna K. P. , Woo Patrick C. Y. 
Published: July 2014

ABSTRACT

We report the complete mitochondrial genome sequence of
Lichtheimia ramosa
(syn.
Lichtheimia hongkongensis
), the first complete mitochondrial DNA sequence of the genus
Lichtheimia
. This 31.8-kb mitochondrial genome encodes 11 subunits of respiratory chain complexes, 3 ATP synthase subunits, 25 tRNAs, and small and large rRNAs, with the gene order
atp9-cox2-atp6-cox3-cox1-nad2-nad3-cob-nad1-nad6-nad5-nad4l-nad4-atp8.
Draft Genome Sequence of a Novel Streptomyces griseorubens Strain, JSD-1, Active in Carbon and Nitrogen Recycling
Authors: Feng Haiwei , Zhi Yuee , Sun Yujing , Wei Xing , Luo Yanqing , Zhou Pei 
Published: July 2014

ABSTRACT

Streptomyces griseorubens
JSD-1, isolated from compost-treated soil, is able to utilize lignocellulose and nitrate as its sole carbon and nitrogen source for growth. Here, we announce the draft genome map of this actinomycete. The genes participating in lignocellulose and nitrate metabolism were picked out and identified.
Draft Genome Sequence of Beneficial Rice Rhizosphere Isolate Pseudomonas aeruginosa PUPa3
Authors: Uzelac Gordana , Bertani Iris , Kojic Milan , Paszkiewicz Konrad H. , Studholme David J. , Passos da Silva Daniel , Venturi Vittorio 
Published: July 2014

ABSTRACT

Pseudomonas aeruginosa
PUPa3 is a rhizosphere-colonizing and plant growth-promoting strain isolated from the rhizosphere of rice. This strain has, however, been shown to be pathogenic in two nonmammalian infection models. Here we report the draft genome sequence of
P. aeruginosa
PUPa3.
Genome Sequence of Oenococcus oeni OM27, the First Fully Assembled Genome of a Strain Isolated from an Italian Wine
Authors: Lamontanara Antonella , Orrù Luigi , Cattivelli Luigi , Russo Pasquale , Spano Giuseppe , Capozzi Vittorio 
Published: July 2014

ABSTRACT

Oenococcus oeni
OM27 is a strain selected from “Nero di Troia” wine undergoing spontaneous malolactic fermentation. “Nero di Troia” is a wine made from “Uva di Troia” grapes, an autochthonous black grape variety from the Apulian region (south of Italy). In this paper we present a 1.78-Mb assembly of the
O. oeni
OM27 genome, the first fully assembled genome of an
O. oeni
strain from an Italian wine.
Draft Genome Sequence of
Cupriavidus
sp. Strain SK-3, a 4-Chlorobiphenyl- and 4-Clorobenzoic Acid-Degrading Bacterium
Authors: Vilo Claudia , Benedik Michael J. , Ilori Matthew , Dong Qunfeng 
Published: July 2014

ABSTRACT

We report the draft genome sequence of
Cupriavidus
sp. strain SK-3, which can use 4-chlorobiphenyl and 4-clorobenzoic acid as the sole carbon source for growth. The draft genome sequence allowed the study of the polychlorinated biphenyl degradation mechanism and the recharacterization of the strain SK-3 as a
Cupriavidus
species.
Neighbourhood environment factors and the occurrence of injuries in Canadian adolescents: a validation study and exploration of structural confounding
Authors: Vafaei Afshin , Pickett William , Alvarado Beatriz E 
Published: July 2014

ObjectivesSocial sorting mechanisms or analogous selection processes may impose confounding effects in the study of aetiological relationships. Such processes are referred to as structural confounding. If present, certain strata of social factors could hypothetically never be exposed to specific risk factors. This prohibits exchangeability across groups that is needed for meaningful causal inference. The objectives of this study were to: (1) develop and test the reliability and validity of composite scales for the measurement of social capital (SC), socioeconomic status (SES) and built environment (BE) and (2) to explore the possible roles of community level SC, SES and BE factors in studies of the aetiology of youth injury.Setting/participantsA nationally representative sample of over 26 000 Canadian students aged 11–15 years.Measures/analysisScales describing these key factors were developed and validated via exploratory and confirmatory factor analyses. We then used tabular analyses to explore structural confounding in our population.ResultsThe proposed scales all demonstrated good psychometric properties. Despite variations in the number of adolescents across social and environmental strata, no evidence for the presence of structural confounding was detected in our data.ConclusionsRelationships between social capital and the occurrence of injuries in Canadian youth aged 11–16 can potentially be studied without consideration of structural confounding biases. Canada is a suitable place to disentangle the effects of different neighbourhood social and environmental exposures on occurrence of injuries and other outcomes in adolescent populations. Exchangeability is possible across exposure strata and therefore a meaningful multilevel regression analysis is feasible. However, more studies are needed to test the consistency of our findings in other populations and for different outcomes.
Effect of mycophenolate mofetil on progression of interstitial fibrosis and tubular atrophy after kidney transplantation: a retrospective study
Authors: Mihovilović Karlo , Maksimović Bojana , Kocman Branislav , Guštin Denis , Vidas Željko , Bulimbašić Stela , Ljubanović Danica Galešić , Matovinović Mirjana Sabljar , Knotek Mladen 
Published: July 2014

ObjectivesChronic transplant dysfunction after kidney transplantation is a major reason of kidney graft loss and is caused by immunological and non-immunological factors. There is evidence that mycophenolate mofetil (MMF) may exert a positive effect on renal damage in addition to immunosuppression, by its direct antifibrotic properties. The aim of our study was to retrospectively investigate the role of MMF doses on progression of chronic allograft dysfunction and fibrosis and tubular atrophy (IF/TA).SettingRetrospective, cohort study.ParticipantsPatients with kidney transplant in a tertiary care institution. This is a retrospective cohort study that included 79 patients with kidney and kidney–pancreas transplantation. Immunosuppression consisted of anti-interleukin 2 antibody induction, MMF, a calcineurin inhibitor±steroids.Primary outcome measuresAn association of average MMF doses over 1 year post-transplant with progression of interstitial fibrosis (Δci), tubular atrophy (Δct) and estimated-creatinine clearance (eCrcl) at 1 year post-transplant was evaluated using univariate and multivariate analyses.ResultsA higher average MMF dose was significantly independently associated with better eCrcl at 1 year post-transplant (b=0.21±0.1, p=0.04). In multiple regression analysis lower Δci (b=−0.2±0.09, p=0.05) and Δct (b=−0.29±0.1, p=0.02) were independently associated with a greater average MMF dose. There was no correlation between average MMF doses and incidence of acute rejection (p=0.68).ConclusionsA higher average MMF dose over 1 year is associated with better renal function and slower progression of IF/TA, at least partly independent of its immunosuppressive effects.