bluuebunny/crossref_metadata_2025_split huggingface dataset:
Psychometrics evaluation of Charcot‐Marie‐Tooth Neuropathy Score (CMTNSv2) second version, using Rasch analysis
Authors: Sadjadi Reza , Reilly Mary M. , Shy Michael E. , Pareyson Davide , Laura Matilde , Murphy Sinead , Feely Shawna M. E. , Grider Tiffany , Bacon Chelsea , Piscosquito Giuseppe , Calabrese Daniela , Burns Ted M. 
Published: November 2014

AbstractCharcot‐Marie‐Tooth Neuropathy Score second version (CMTNSv2) is a validated clinical outcome measure developed for use in clinical trials to monitor disease impairment and progression in affected CMT patients. Currently, all items of CMTNSv2 have identical contribution to the total score. We used Rasch analysis to further explore psychometric properties of CMTNSv2, and in particular, category response functioning, and their weight on the overall disease progression. Weighted category responses represent a more accurate estimate of actual values measuring disease severity and therefore could potentially be used in improving the current version. Copyright © 2014 John Wiley & Sons, Ltd.
Body Image in Women with Primary and Secondary Provoked Vestibulodynia: A Controlled Study
Authors: Maillé Delphine L. , Bergeron Sophie , Lambert Bernard 
Published: November 2014

Abstract

Introduction
Provoked vestibulodynia (PVD) is a women's genito-pelvic pain condition associated with psychosexual impairments, including depression. Body image (BI) has been found to be different in women with primary (PVD1) and secondary (PVD2) PVD. No controlled study has compared BI in women with PVD1 and PVD2 and investigated its associations with sexual satisfaction, sexual function, and pain.

Aims
The aims of this study were to (i) compare BI in women with PVD1, PVD2, and asymptomatic controls and (ii) to examine associations between BI and sexual satisfaction, sexual function, and pain during intercourse in women with PVD.

Methods
Fifty-seven women (20 with PVD1, 19 with PVD2, and 18 controls) completed measures of BI, sexual satisfaction, sexual function, pain during intercourse, and depression.

Main Outcome Measures
The main outcome measures were (i) Global Measure of Sexual Satisfaction Scale, (ii) Female Sexual Function Index, and (iii) pain numerical rating scale.

Results
Controlling for depression, women with PVD1 reported more body exposure anxiety during sexual activities than women with PVD2 and controls F(2,51) = 4.23, P = 0.02. For women with PVD, more negative BI during sexual activities was associated with lower sexual satisfaction (β = −0.45, P = 0.02) and function (β = −0.39, P = 0.04) and higher pain during intercourse (β = 0.59, P = 0.004). More positive body esteem was associated with higher sexual function (β = 0.34, P = 0.05).

Conclusions
Findings suggest that women with PVD1 present more body exposure anxiety during sexual activities than women with PVD2 and asymptomatic women. Body esteem and general attitudes toward women's genitalia were not significantly different between groups. Higher body exposure anxiety during sexual activities was associated with poorer sexual outcomes in women with PVD. Further studies assessing interventions targeting BI during sexual activities in this population are needed, as improving BI during sexual interactions may enhance sexual outcomes in women with PVD.
Chores and sense of self: gendered understandings of voices of older married women with dementia
Authors: Hellström Ingrid , Eriksson Henrik , Sandberg Jonas 
Published: November 2014

BackgroundMarital relationships in dementia are forged between the person with dementia and the care partner, and such relationships have an impact on the way in which dementia is understood and experienced. The everyday work that underpins the relationship is usually divided between spouses and based on traditional divisions of household chores.Aims and objectivesThe aim was to describe how older women with dementia express the importance of their homes and their chores in everyday life.MethodsSeven women with dementia, who were cohabiting with their husbands, were interviewed on up to five occasions at home during a five‐to‐six‐year period on the following themes: the home, their dementia illness, everyday life, their relationships with their husbands and dignity and autonomy.ResultsThe qualitative analysis showed three different patterns in the women's narratives: keeping the core of the self through the home, keeping the self through polarising division of labour and keeping the self through (re‐) negotiations of responsibilities. The feeling of one's home and home‐related chores is an essential way to express who you are.ConclusionThe women stated that household chores are the centre of their lives despite their dementia disease and that the home, even though it shrinks, still makes the women see themselves as an important person, namely the ‘competent wife’.Implications for practiceNurses need to be aware that ‘doing gender’ may be a means of preserving personhood as well as of sustaining couplehood in dementia.
Do we look happy? Perceptions of romantic relationship quality on Facebook
Authors: EMERY LYDIA F. , MUISE AMY , ALPERT ELIZABETH , LE BENJAMIN 
Published: November 2014

AbstractDoes the information that people share about their romantic relationships on Facebook influence how other people perceive their relationship quality? Across 2 studies, people's accuracy at inferring others' relationship quality based on their Facebook profiles and how Facebook profiles influenced others' judgments of people's relationship quality and likability were investigated. Perceived relationship quality corresponded to self‐reported relationship quality (Study 1), and people with more visible relationships were perceived as having greater relationship satisfaction and commitment (Studies 1 and 2) and being more likable (Study 2). High disclosure about the relationship predicted greater perceived relationship quality but lower likability (Study 2). These findings illuminate how sharing information about one's relationship influences other people's impressions of the individual and the relationship.
Optimized Surface Treatments and Manufacturing Techniques for Components of Sodium Cooled Fast Reactors
Authors: Horowitz Emmanuel 
Published: November 2014

The French sodium prototype ASTRID should work from 2026; it must satisfy the requirements of “Generation IV” with regard to lifetime, safety and efficiency; means are sought in order to optimize the ageing behavior of components.
Coatings that seem favorable concerning the fuel claddings and wrappers of ASTRID must be further checked.
Other surface treatments are considered for less irradiated components, in order to improve the residual stresses near the surface of metallic pieces; if the stresses get compressive, these treated components get less brittle and “stress corrosion cracking” (SCC) may be avoided.
French and Chinese establishments have developed “ultrasonic shot peening” using stainless steel balls or needles. Concurrently, US and Japanese companies have carried out “water jet peening” and nanopulse “Laser Peening” treatments; These “laser peening” treatments have already enhanced efficiently SCC and fatigue behavior of both AISI 316 pieces and 12%Cr ferrochromium blades.
Laser treatments with lower power requirements improve also the behaviour of surfaces with regard to SCC, roughness or pitting corrosion.
Passivating chemical techniques are required too in order to stabilize the protective chromine layer at the surface of metallic pieces. Peening and passivating techniques should be combined in the future.
Detailed Observation of Seawater Precipitation With 5×5 Mockup Fuel Bundle Based on X-Ray Computated Tomography Technique
Authors: Kojo Retsu , Hotta Akitoshi , Furuya Masahiro , Akiba Miyuki , Hoshi Harutaka 
Published: November 2014

In order to conform to the new regulatory standard in Japan, seawater is regarded as the alternative water source both for BWRs (Boiling water reactors) and PWRs (Pressurized water reactors). For preventing further accident evolutions occurred in Fukushima Daiichi nuclear power plants, seawater was injected into the reactors for more than one week. With long-term seawater injection, sea salt compositions are condensed and many of them will precipitate when saturated concentrations are exceeded. Unlike corrosion issues, impacts of sea salt precipitation on the heat removal has not been studied widely in the past because it has not been regarded as the alternative water source before Fukushima Daiichi nuclear power plants accident.
The existent knowledge base of boric acid precipitation under LOCA conditions was studied. Based on the existent study on impacts of boric acid precipitation under LOCA conditions, the experimental project of seawater consisting of four experiments was proposed.
The existent database of physical properties, such as viscosity, of seawater is rather poor under severe accident conditions. In addition, it is likely that boric acid will be injected with water or seawater to prevent re-criticality. It is known that physical properties of boric acid vary widely under high temperature conditions. Measurement of viscosity of the seawater-boric acid mixture was conducted in high temperature using a rotational viscometer. Under conditions equivalent to the estimated bulk coolant conditions under a long-term cooling phase of severe accidents. In this range, no obvious change of viscosity is expected.
Then a detailed structure of seawater precipitates was observed using a mockup fuel bundle with 5×5 in the square lattice and 500 mm in the length. Images of precipitates were taken using the X-ray CT. The water level, the concentration of sea salt and the heat flux are employed as experimental parameters. The heat flux, bubble stirring in downstream of spacers and heat loss by a non-heated channel box were identified as influential factors to local and overall precipitates in fuel bundles.
Heat Removal Capability of Core-Catcher in Consideration of Transformation of Cooling Channel
Authors: Kurita Tomohisa , Komuro Mitsuo , Suzuki Ryo , Yamada Masato , Tahara Mika , Yanagisawa Hiromasa 
Published: November 2014

It is necessary to stabilize high temperature molten core in a severe accident for long time without electrical power. The core-catcher is to be installed at the bottom of the lower drywell in order to settle the molten core flowing down from a reactor vessel.
Toshiba’s core-catcher system consists of a round basin made up of inclined cooling channels to get natural circulation of the flooding water. So it can cover all pedestal floor and can work in passive manner.
We have been confirming an applicability of the core-catcher to actual plants. We have conducted full scaled tests with a unique cooling channel which has inclined rectangular flow section and changing the section area along flow direction in several conditions to evaluate the influence of the parameters on the natural circulation and heat removal capability. The test results showed good heat removal performance with nucleate boiling.
However, we should consider a transformation of the cooling channel, for example, by the falling corium. So we calculate the assumed transformation of the cooling channel and conduct natural circulation tests with obstruction in the cooling channel. We confirm that natural circulation flow is stably continues and the cooling channel can remove prescribed heat, even if a flow area have got narrow locally.
Japanese Codes and Standards for Dual Purpose Metal Casks for Spent Nuclear Fuel
Authors: Saegusa Toshiari , Hirose Makoto , Irie Norikazu , Shimizu Masashi 
Published: November 2014

The first Japanese spent fuel interim storage facility away from a reactor site is about to be commissioned in Mutsu City, Aomori Prefecture. In designing, licensing and construction of the Dual Purpose Casks (DPCs, for transport and storage) for this facility, codes and standards established by the Atomic Energy Society of Japan (AESJ) and by the Japan Society of Mechanical Engineers (JSME) have been applied.
The AESJ established the first standard for DPCs as “Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facilities” in 2002 (later revised in 2010). The standard provides the design requirements to maintain the basic safety functions of DPCs, namely containment, heat removal, shielding, criticality prevention and the structural integrity of the cask itself and of the spent fuel cladding during transport and storage. Inspection methods and criteria to ensure maintenance of the basic safety functions and structural integrity over every stage of operations involving DPCs including pre-shipment after storage are prescribed as well. The structural integrity criteria for major DPC components refer to the rules provided by the JSME.
JSME completed the structural design and construction code (the Code) for DPCs as “Rules on Transport/Storage Packagings for Spent Nuclear Fuel” in 2001 (later revised in 2007). Currently, the scope of the rules cover the Containment Vessel, Basket, Trunnions and Intermediate Shell as major components of DPCs. Rules for these components are based on those for components of nuclear power plants (NPP) with similar safety functions, but special considerations based on their shapes, loading types and required functions are added.
The Code has differences from that for NPP components with considerations to DPC characteristics;
- The primary stress and the secondary stress generated in Containment Vessels shall be evaluated under Service Conditions A to D (from ASME Sec III, Div.1).
- Stress generated in the seal region lid bolts of Containment Vessels shall not exceed yield strength under Service Conditions A to D in order to maintain the containment function.
- Fatigue analysis on Baskets is not required, and Trunnions can be designed only for Service Conditions A and B with special stress limits consistent with conventional assessment methods for transport packages.
- Stress limits for earthquakes during storage are specified.
- Ductile cast iron with special fracture toughness requirements can be used as a material for Containment Vessels.
DPC specific considerations in standards and rules will be focused on in this paper. Additionally, comparison with the ASME Code will be discussed.
Development of Steam Generator Model Using RELAP5 to Simulate Two-Phase Natural Circulation for Pressurized Water Reactor
Authors: Miyata Shinya , Kamohara Satoru , Sakuma Wataru , Nishi Hiroaki 
Published: November 2014

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are.
Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.
Modeling of Cladding Thermal Evolution Along Cask Storage
Authors: Penalva Jaime E. , Feria Francisco , Herranz Luis E. 
Published: November 2014

One of the safety requirements in dry storage of spent fuel is to ensure the cladding integrity. In this regard, the understanding of the cladding mechanical performance along the storage period is indispensable, both to analyse the failure probability and to characterize the state of the cladding so that fuel management is conducted with accurate knowledge of the material conditions. The main interest is focused on cladding degrading mechanisms as creep and hydrogen related (e.g. hydrides embrittlement), which are strongly influenced by temperature. Therefore, cladding thermal characterization along dry storage is an important element to predict fuel rod mechanical performance.
Cladding temperature decay models found in the literature are fuel burnup independent and they cannot be applied to storage periods longer than some decades. The goal of this work is to develop a simplified model of cladding temperature as a function of burnup that spans up to 300 years of cask storage. To do so, a methodology is established based on FLUENT steady state calculations fed by heat decay data found in the literature for different burnups (33–63 MWd/kgU). From the results, a temperature correlation as a function of burnup and out-of-reactor time has been derived. It shows an average relative error less than 2% with respect FLUENT calculations.
Finally, significance of having an accurate thermal characterization of the fuel rod has been highlighted by comparing fuel rod thermo-mechanics based on the derived correlation and the one resulting from using a correlation developed by EPRI.
Implications of Applying Semi-Scale Pump Test Data to Two-Phase RCP Performance Modeling for PWR LOCA Analyses
Authors: Jacob Mathew C. , Coon Michael T. , Blaisdell John A. , Espinosa Ruben J. 
Published: November 2014

Emergency Core Cooling System (ECCS) analyses using Loss of Coolant Accident (LOCA) codes utilize two-phase Reactor Coolant Pump (RCP) performance models formulated on the basis of data from tests conducted on the Semi-scale pump (Reference 1) operating at 60 Hz frequency. In some PWRs, the RCPs operate at a frequency of 50 Hz. This paper presents the results of an evaluation performed to determine the applicability of RCP two-phase performance models developed on the basis of data from the Semi-scale tests for analyzing ECCS performance of new generation PWRs. The evaluation addressed two major issues: (1) the applicability of the two-phase RCP performance model developed using the data from the Semi-scale pump tests (Reference 1) for full scale Pressurized Water Reactor (PWR) LOCA simulations, and (2) the relevance of the two-phase RCP performance model developed on the basis of test data for the Semi-scale pumps running at 60 Hz frequency to PWR RCPs running at 50 Hz frequency with higher specific speeds.
Reviews of pump performance test data available in the open literature identified two-phase performance data appropriate for use in substantiating the validity of current PWR pump performance models. These data supported the conclusion that the two-phase head performance degradation for the Semi-scale Mod-1 pump is conservative compared to the two-phase pump performance data generated from testing of pumps representative of full scale PWR RCPs. A review of ECCS analyses results available in the literature determined that the use of the current RCP two-phase performance model (developed using the Semi-scale Mod-1 pump test data) for a typical PWR plant resulted in about a 100 °F increase in the Peak Clad Temperature (PCT) for a Large Break LOCA (LBLOCA) in comparison to the PCTs calculated using the two-phase pump performance model developed on the basis of test data for pumps representative of full scale PWR RCPs.
It was determined from the current study that the frequency (50 Hz vs. 60 Hz) of the electrical power that drives the pump motor is not of much consequence for two-phase RCP performance modeling, since (1) the RCP performance model is characterized via normalized pump performance parameters, and (2) for the LBLOCA analysis of interest, the RCPs are assumed to lose power at the start of the event.
Modelling of Water Ingress in to Vacuum Vessel
Authors: Kaliatka Tadas , Ušpuras Eugenijus , Kaliatka Algirdas 
Published: November 2014

An event of water coolant ingress into vacuum vessel is one of the most important events leading to severe consequences in nuclear fusion reactors. The ingress of coolant to the vacuum vessel could appear due to coolant pipe rupture of in-vessel components. Vacuum vessel could not withstand the high pressure inside. Pressure increase in vacuum vessel is due to water evaporation because of pressure difference and water contact with high temperature plasma facing components. If pressure in vacuum vessel is too high — safety valve opens and the steam containing activated dust will be transferred form the vacuum vessel to the environment. Therefore, it is important to understand the thermo hydraulic processes in vacuum vessel during the ingress of coolant event (ICE).
There are few experimental investigations performed, modeling of ICE. In this article ingress of coolant event in vacuum vessel was modeled using RELAP5 code. RELAP5 is a “best estimate” system code suitable for the thermo-hydraulic analysis of all transients and postulated accidents in nuclear fission, light water reactor systems, including both large and small-break loss-of-coolant accidents as well as the full range of operational transients. The use of RELAP5 code for the accident analysis in nuclear fusion reactors allows to perform integral analysis of thermal-hydraulic processes in the cooling system and vacuum vessel. The comparisons of calculation results and experimental data showed that with some limitations the RELAP5 program could be used for the analysis of the thermal hydraulic processes in the vacuum vessel during ICE.
Shipment of Spent Nuclear Fuel From the Czech Republic for Reprocessing
Authors: Podlaha Josef 
Published: November 2014

After more than 50 years of operation of the research reactor operated by ÚJV Řež, a. s. (ÚJV) or the Research Centre Řež, as the case may be, a large amount of spent nuclear fuel (SNF) of Russian origin has accumulated.
In 2005, ÚJV joined the Russian Research Reactor Fuel Return (RRRFR) program under the US-Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF) using the ŠKODA VPVR/M transport packaging system (TPS). Two SNF shipments from ÚJV were carried out in 2007 and 2013. After the shipments were completed, only low-enriched nuclear fuel with a maximum enrichment below 20% of 235U remained on the territory of the Czech Republic.
ÚJV also participates in shipments of SNF from other countries. The services of ÚJV comprise mainly ŠKODA VPVR/M TPS leasing, technical oversight and expertise during cask handling, SNF loading and cask closing and sealing. Up to now, ÚJV has participated in thirteen shipments of SNF from eight countries; one shipment is currently being prepared.
High-level radioactive waste (HLW) will be generated from SNF reprocessing. The vitrified HLW will be returned to the Czech Republic as stated in the Russian-Czech Intergovernmental Agreement on Co-operation in Nuclear Energy. The return of the waste represents very complex and complicated work, technically, legally and contractually.